Operational Radiation Safety

NCRP Report No. 142

Operational Radiation SafetyProgram for Astronauts in Low-Earth Orbit: A Basic Framework

Recommendations of theNATIONAL COUNCIL ON RADIATIONPROTECTION AND MEASUREMENTS

Issued November 30, 2002

National Council on Radiation Protection and Measurements7910 Woodmont Avenue, Suite 400 / Bethesda, Maryland 20814

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National Council on Radiation Protection and Measurements.

Operational radiation safety program for astronauts in low-earth orbit : a basic

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‘‘Issued November 2002.’’

Includes bibliographical references and index.

ISBN 0-929600-75-4

1. Extraterrestrial radiation—Safety measures. 2. Astronauts—Health and

hygiene. 3. Space medicine. I. Title. II. Series

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Preface

This Report was developed under the auspices of Scientific Com-

mittee 46, the National Council on Radiation Protection and

Measurements (NCRP) program area committee concerned with

operational safety. This Report addresses the operational radiation

safety program for astronauts working in low-Earth orbit, with par-

ticular attention to the radiation dosimetry needed, the radiation

exposure information that should be recorded, and ways to imple-

ment the radiation protection principle of ‘‘as low as reasonably

achievable’’ for activities involving the Space Shuttle and the Inter-

national Space Station. It is a companion document to NCRP Report

No. 132, Radiation Protection Guidance for Activities in Low-Earth

Orbit, published in December 2000, and provides advice on imple-

mentation of NCRP Report No. 132 guidance.

This work was performed at the request of the National Aeronau-

tics and Space Administration (NASA) and NCRP gratefully

acknowledges NASA’s support. The Scientific Committee that pre-

pared this Report also benefited from a series of briefings from NASA

staff at the Committee’s first meeting in December 1999 in Houston.

This Report was prepared by Scientific Committee 46-15 on Opera-

tional Radiation Safety Program for Astronauts. Serving on Scien-

tific Committee 46-15 were:

Richard J. Vetter, Chairman

Mayo Clinic

Rochester, Minnesota

Members

Ellen S. Baker David T. Bartlett

National Aeronautics and National Radiological

Space Administration Protection Board

Lyndon B. Johnson Space Chilton, Oxon, United

Center Kingdom

Houston, Texas

Thomas B. Borak Susan M. Langhorst

Colorado State University Washington University in

Fort Collins, Colorado St. Louis School of Medicine

St. Louis, Missouri

iii

iv / PREFACE

Stephen W.S. McKeever Jack Miller

Oklahoma State University Lawrence Berkeley National

Stillwater, Oklahoma Laboratory

Berkeley, California

R. Julian Preston John W. Wilson

U.S. Environmental Protection National Aeronautics and

Agency Space Administration

Research Triangle Park, Langley Research Center

North Carolina Hampton, Virginia

Advisor

Charles B. Meinhold

National Council on Radiation Protection and Measurements

Bethesda, Maryland

NCRP Secretariat

Marvin Rosenstein, Consultant

Cindy L. O’Brien, Managing Editor

The Council wishes to express its appreciation to the Committee

members for the time and effort devoted to the preparation of this

Report.

Thomas S. Tenforde

President

Contents

Preface ........................................................................................ iii

1. Summary and Recommendations ................................... 1

1.1 Components of an Operational Radiation Safety

Program ......................................................................... 1

1.2 Team Management in the Radiation Safety Program 2

1.3 Sources of Radiation in Space ...................................... 3

1.4 Dose Limits for Astronauts ........................................... 3

1.5 Sources of Exposure Included in the Dose Limits

for Astronauts ................................................................ 4

1.6 Types of Radiation to be Assessed ............................... 5

1.7 Approach to Dose Assessment for Astronauts ........... 6

1.8 Operational Radiation Monitoring ............................... 7

1.8.1 Area Monitoring .................................................. 7

1.8.2 Personal Dosimetry ............................................. 8

1.8.3 Calibration ........................................................... 10

1.9 Biodosimetry .................................................................. 11

1.10 Immediate Dose Management and ‘‘As Low As

Reasonably Achievable’’ ................................................ 11

1.11 Radiation Safety Training ............................................ 12

1.12 Dosimetry Record .......................................................... 13

2. Objectives of the Operational Radiation Safety

Program for Astronauts .................................................... 15

2.1 The Low-Earth Orbit Program ...................................... 15

2.2 Dose Limits for Astronauts ........................................... 17

2.2.1 Deterministic Limits ............................................. 18

2.2.2 Stochastic Limits .................................................. 18

2.3 Operational Radiation Protection Considerations ........ 22

3. Current Management of Astronaut Radiation Safety

Program ................................................................................. 24

3.1 Flight Rules for Management of Dose ........................... 24

3.2 Biomedical Research ....................................................... 26

3.3 Individuals Involved in Management of Dose ............... 26

3.3.1 Astronaut ............................................................. 26

3.3.2 Flight Director ...................................................... 26

3.3.3 Flight Surgeon ...................................................... 27

3.3.4 Radiation Health Officer ..................................... 27

3.3.5 Space Radiation Analysis Group ......................... 28

v

vi / CONTENTS

4. Radiation Environment in Low-Earth Orbit ............... 294.1 Trapped Electrons (0.5 to 6 MeV; �0.2 keV �m�1) ...... 304.2 Reentrant and Splash Albedo Electrons (1 MeV to

�1 GeV; 0.2 to �3 keV �m�1) ........................................ 314.3 Trapped Protons (�10 MeV; �5 keV �m�1) .................. 314.4 Trapped and Solar Protons and Light Nuclear

Particles (10 to 400 MeV; 0.3 to 5 keV �m�1) .................. 314.5 Galactic Cosmic Radiation Ions and High-Energy

Secondary Fragments (�50 MeV n�1; Z � 1; 1 to

1,000 keV �m�1) .............................................................. 324.6 Charged Target Fragments (�10 MeV n�1; 2 to

1,200 keV �m�1) .............................................................. 324.7 Neutrons (0.1 to 500 MeV) ............................................. 334.8 Summary for Particle Types ........................................... 33

5. Approach to Dose Assessment for Astronauts ............. 355.1 Exterior Exposure Field ................................................. 38

5.1.1 Radiation Environment Models ........................... 385.1.2 Spacecraft External Measurements .................... 39

5.2 Interior Exposure Field ................................................... 395.2.1 Shielding Models for the Space Shuttle,

International Space Station, and Space Suits .... 395.2.2 Spacecraft Internal Measurements ..................... 40

5.3 Tissue Exposure Fields ................................................... 415.3.1 Human Shielding Models ..................................... 415.3.2 Occupancy Factors ................................................ 41

6. Data Collection and Interpretation for Dose

Assessment ........................................................................... 426.1 Introduction .................................................................... 426.2 Dose Quantities to be Determined ................................ 436.3 Proposed Measurement Package .................................... 44

6.3.1 General Discussion ............................................... 446.3.2 Proposed Measurement Package: Active

Devices ................................................................... 466.3.2.1 Tissue Equivalent Proportional

Counters ................................................... 476.3.2.2 Solid-State Detectors ............................... 476.3.2.3 Active Electronic Personal Dosimeters .... 486.3.2.4 Active Detectors for Electrons ................ 48

6.3.3 Proposed Measurement Package: Passive

Devices .................................................................. 496.3.3.1 Low Linear Energy Transfer

Dosimetry: Thermoluminescent

Dosimeters ............................................... 506.3.3.2 Direct Ion Storage Dosimeters ............... 51

CONTENTS / vii

6.3.3.3 Neutron and High Atomic Number,

High-Energy Particle Dosimetry:

Plastic Nuclear Track Detectors ............ 51

6.3.3.4 Use of Thermoluminescent Dosimeters

and Plastic Nuclear Track Detectors to

Estimate Effective Dose .......................... 52

6.3.3.5 Superheated Drop/Bubble Dosimeters ..... 52

6.3.4 Recommendations for Measurement Packages .. 53

6.3.4.1 Recommendations for Area Monitoring ... 53

6.3.4.2 Recommendations for Personal

Dosimetry ................................................ 54

6.4 Accuracy, Performance Testing, and Calibration ........ 55

6.4.1 Operational Radiation Protection Requirements

on Accuracy of Dose Measurements .................... 55

6.4.1.1 Recommendations of ICRP and ICRU ... 55

6.4.1.2 General Requirements ........................... 56

6.4.2 Tests of Instrument and Dosimeter

Performance .......................................................... 56

6.4.3 Calibration ............................................................ 57

7. Role of Biodosimetry in Dose Assessment .................. 59

7.1 Electron Spin Resonance ................................................ 59

7.2 Biochemical Indicators ................................................... 60

7.3 Erythrocytes with Transferrin Receptors ...................... 60

7.4 Gene Mutation Assays ................................................... 60

7.5 Cytogenetic Alterations ................................................... 61

7.5.1 Micronuclei ............................................................ 61

7.5.2 Acentric Fragments in Prematurely Condensed

Chromosomes ........................................................ 62

7.5.3 Chromosomal Aberrations .................................. 62

7.6 Summary for Methods .................................................... 63

7.7 Current NASA (Lyndon B. Johnson Space Center)

Methods ........................................................................... 63

7.8 Recommendations and Future Considerations ............. 64

8. Recommended Management of Astronaut Radiation

Safety Program .................................................................... 66

8.1 Components of a Low-Earth Orbit Operational

Radiation Safety Program ............................................. 66

8.2 Radiation Protection Principles Applied to Low-Earth

Orbit Missions ................................................................ 67

8.3 Sources of Exposure Included in NCRP Dose Limits

for Astronauts ................................................................. 68

viii / CONTENTS

8.4 Immediate Dose Management and Basic ‘‘As Low As

Reasonably Achievable’’ Concepts .................................. 69

8.4.1 Immediate Dose Management Issues ................. 70

8.4.2 Basic ‘‘As Low As Reasonably Achievable’’

Concepts ................................................................ 70

8.4.3 Considerations for Spacecraft and Space Suit

Design .................................................................... 70

8.4.4 Considerations for Preflight Planning ................ 71

8.4.5 Considerations for Continuous In-Flight

Review ................................................................... 71

8.4.6 Considerations for Postflight Review .................. 72

8.5 Radiation Safety Training for NASA Personnel ........... 72

8.5.1 Astronauts ............................................................. 73

8.5.2 Flight Directors ..................................................... 74

8.5.3 Flight Surgeons ..................................................... 75

8.5.4 Radiation Health Officer and Radiation Safety

Support Groups .................................................... 75

8.5.5 Other Supporting Specialists ............................... 76

9. Radiation Safety Records ................................................. 77

9.1 Content of Records .......................................................... 78

9.1.1 Records of Career Doses ....................................... 80

9.1.2 Records of Prospective and Retrospective

Studies .................................................................. 81

9.2 Continuity of Records Over Time ................................... 82

9.2.1 Existing Records and Unanalyzed Data Files .... 82

9.2.2 Future Adjustment of Records ............................. 82

9.3 Retention of Records ....................................................... 83

Appendix A. Time-Dependent Variations in the Space

Radiation Environment and the Need for Active

Real-Time Monitoring ........................................................ 84

A.1 Background ..................................................................... 84

A.2 Types of Active Detectors Used in Space ...................... 86

A.2.1 Tissue Equivalent Proportional Counters .......... 86

A.2.2 Semiconductor Detectors ..................................... 91

A.2.3 Cerenkov Counters ............................................. 93

A.2.4 Ionization Chambers ........................................... 93

Appendix B. Computational Methods ................................ 95

Appendix C. Thermoluminescence Dosimetry

Materials ............................................................................... 99

CONTENTS / ix

C.1 Lithium Fluoride, Doped with Magnesium and

Titanium ........................................................................ 99

C.2 Calcium Fluoride, Doped with Thulium ..................... 105

C.3 Lithium Fluoride, Doped with Magnesium, Copper

and Phosphorus ............................................................ 107

C.4 Aluminum Oxide, Doped with Carbon ........................ 108

C.5 Evaluation of Absorbed Dose and Dose Equivalent ..... 110

C.5.1 Equivalent Gamma-Ray Dose ........................... 110

C.5.2 Mean, or Effective, Linear Energy Transfer .... 111

C.5.3 Dose Equivalent ................................................. 112

Appendix D. Plastic Nuclear Track Detectors ................. 115

D.1 Description of Method .................................................. 115

D.2 Neutron Dosimetry ..................................................... 116

D.3 Dosimetry of Cosmic Radiation Fields ....................... 117

Glossary .................................................................................... 122

Acronyms, Abbreviations, and Main Symbols .................. 126

References ................................................................................ 128

The NCRP .................................................................................. 141

NCRP Publications .................................................................. 150

Index ........................................................................................... 160

1. Summary andRecommendations

Astronauts are living and working for extended periods in low-

Earth orbit (LEO) during Space Shuttle missions and construction,

maintenance and operation of the International Space Station (ISS).

The radiation environment they encounter in space is complex, with

unique high-LET (linear energy transfer) and high-energy compo-

nents, as distinct from the predominately low-LET and low-energy

radiation environments encountered by most radiation workers on

Earth. The primary purpose of an operational radiation safety pro-

gram for astronauts working in LEO is to assess and control the

radiation exposure of individual astronauts commensurate with mis-

sion tasks and the prevailing radiation conditions in LEO.

1.1 Components of an Operational Radiation

Safety Program

The main components of an operational radiation safety program

designed to implement the principles of dose limitation and ALARA

(as low as reasonably achievable) for astronauts working in LEO are:

● to facilitate actions, both in advance of a mission and in-flight,

that respond to space radiation conditions or mission decisions

that significantly affect the level of radiation exposure to the

astronauts, and radiation protection decisions that significantly

influence the conduct of the mission;

● to collect and record data to assess astronaut doses for individual

mission and cumulative career records; and

● to identify, plan and carry out practical ALARA actions to avoid

unnecessary levels of radiation exposure.

Recommendation 1: National Aeronautics and Space Admin-

istration (NASA) management should implement and main-

tain an effective radiation safety program with the following

features: clear definition of the goals of the program, state-

ment of the organization’s commitment to the application

1

2 / 1. SUMMARY AND RECOMMENDATIONS

of the ALARA principle, statement of management’s com-

mitment to provide adequate budgetary support for the

program, and periodic review of the overall program

performance.

Recommendation 2: NASA management should clearly assign

responsibility for ensuring the translation of radiation pro-

tection strategies and instrumentation from design and

development through engineering, preparation for flight,

and use in orbit.

1.2 Team Management in the Radiation Safety Program

Management of the radiation safety program in LEO is a team

effort, involving the astronauts, the flight director, the flight surgeon,

the radiation health officer (RHO), and the Space Radiation Analysis

Group (SRAG). Typically, astronauts do not play an active role in

decision making and policy regarding radiation protection issues.

Instead, the flight director and flight surgeon direct their radiation

protection actions, with the help of radiation experts. The current

roles of these individuals are noted below.

Usually, one or two astronauts with medical backgrounds repre-

sent the U.S. Astronaut Office’s position regarding radiation protec-

tion at the various meetings and committees. If radiation exposure

is expected to be more than minimal, the affected astronauts may

participate more actively concerning their particular flight.

The flight director is the final decision maker in the Mission Con-

trol Center with regard to all aspects of a mission, and the flight

director relies heavily on the radiation team and flight surgeon when

decisions regarding radiation protection issues need to be made.

The flight surgeon is responsible for the crew’s health and safety

during all aspects of flight, and briefs the crew pre- and postflight

regarding radiation protection issues and personal radiation dose.

The RHO is involved in development and design of radiation pro-

tection strategies and provides recommendations to minimize crew

exposures during mission planning and during space missions,

tracks crew exposures against career limits, and provides risk inter-

pretation for acute exposures.

SRAG consists of NASA radiation specialists who are responsible

for promoting ALARA in development and design of radiation protec-

tion strategies and ensuring compliance with ALARA procedures.

During a mission, SRAG provides an interface to update mission

significant radiation events, particularly when transient events in

1.4 DOSE LIMITS FOR ASTRONAUTS / 3

the space radiation environment produce a potential for high doses

to the astronauts.

Recommendation 3: NASA’s operational radiation safety pro-

gram for LEO should have clearly defined responsibilities

given to an individual or group to ensure overall implementa-

tion of the program. This individual or group may include

individuals from those listed above, and/or members of NASA

management who are able to work across all levels of opera-

tion to ensure radiation safety actions are considered for

implementation.

1.3 Sources of Radiation in Space

The predominant sources of radiation in LEO are galactic cosmic

radiation (GCR) (high-energy protons, helium ions, and heavy ions

of extra-solar origin); solar particle events (SPE) (primarily medium-

energy protons of solar origin); the radiation belts outside Earth’s

atmosphere (high-energy protons and electrons trapped in Earth’s

magnetic field); and scattering from Earth’s atmosphere (albedo neu-

trons, electrons and protons). There is a real potential that high

transient radiation doses to astronauts will occur occasionally during

construction and operation of ISS, particularly from SPEs and rela-

tivistic electrons from the outer radiation belt.

1.4 Dose Limits for Astronauts

The current recommended dose limits for astronauts in LEO were

developed in NCRP Report No. 132 (NCRP, 2000a). The limits for

bone marrow, lens of the eye, and skin are for protection against

deterministic effects. The career limits are for protection against

delayed stochastic effects and are based on a lifetime excess risk of

cancer mortality of three percent.

The following formulations and terminology are used in this Report

for the dose-limit quantities for space activities:

● The dose limits for the relevant organs or tissues for determinis-

tic effects are expressed in terms of gray equivalent, where gray

equivalent is the mean absorbed dose in an organ or tissue

modified by a recommended value, for radiation protection pur-

poses, of the relative biological effectiveness of a given particle

type, as given in NCRP Report No. 132 (NCRP, 2000a). The


recommended values and dose limits for deterministic effects

are given in Tables 2.2 and 2.3, respectively. A conventional

notation GT � RiDT is proposed for gray equivalent and used in

this Report. GT is gray equivalent, Ri is the recommended value

for relative biological effectiveness of a given particle type i, and

DT is the mean absorbed dose in an organ or tissue.

● The career limits for delayed stochastic effects are expressed in

terms of effective dose (E), where:

E � �T

wTHT . (1.1)

HT is the equivalent dose and wT is the tissue weighting factor.

The career limits are given in Table 2.4 and are specified by

gender and age.

● For the complex mixtures of high- and low-LET radiations exper-

ienced in LEO, the practice in the space radiation protection com-

munity is to obtain point values of absorbed dose (D) and dose

equivalent (H) {using D and the quality factor relationship as a

function of LET [Q(L)]}. The point quantities are then averaged

over the organ or tissue of interest by means of computational

models to obtain the organ dose equivalent (ICRU, 1993), which

has been assigned the symbol HT in this Report. This practice

permits more complete consideration of the Q(L) relationship for

these complex radiation environments. The currently recom-

mended Q(L) relationship is given in NCRP Report No. 116

(NCRP, 1993) and is shown in Equation 2.5. For space radiations,

NCRP Report No. 132 (NCRP, 2000a) adopted HT, for operational

radiation protection purposes, as an acceptable approximation for

HT for stochastic effects.

Recommendation 4: For the operational radiation safety pro-

gram in LEO, organ dose equivalent (HT) should be used as

the approximation for equivalent dose (HT).

1.5 Sources of Exposure Included in the Dose Limits

for Astronauts

Recommendation 5: The dose limits for astronauts should

include the cumulative dose from space flight, the dose asso-

ciated with mission-related aviation activities (excluding

commercial flights), the dose from biomedical research con-

ducted as part of the astronaut’s mission duties, and any

1.6 TYPES OF RADIATION TO BE ASSESSED / 5

other occupational doses including any received prior to

work as an astronaut.

The dose limits do not include normal background radiation on

Earth or radiation dose received from diagnostic and therapeutic

medical procedures conducted as part of the astronaut’s overall

health care. In addition, previous medical radiation doses, from diag-

nostic and therapeutic medical procedures, are assumed to have

provided the individual a greater benefit than the risk associated

with the doses and should not be used in determining qualification

for future occupational exposure.

1.6 Types of Radiation to be Assessed

The radiation environment external to a spacecraft in LEO con-

sists of electrons, positrons, neutrons, protons and heavier nuclei

[up to particle charge (Z) � 92]. Energies range from a few electron

volts for trapped electrons and albedo neutrons to in excess of

1014 MeV for GCR ions. Most of the electrons will not penetrate the

wall of the spacecraft, but could penetrate the space suits worn

during extravehicular activity (EVA), resulting in doses to the skin

and eyes. Nuclear interactions of neutrons, protons and heavier

nuclei with spacecraft, space suits, Earth’s atmosphere, and the

human body produce secondary particles, which add to the radiation

field. Radiation monitoring strategies vary according to particle

charge, particle type, energy, and measurement location. The envi-

ronment can be classified according to particle types and energies

and where the measurements are to be made (i.e., outside the space-

craft, inside the spacecraft, and inside EVA suits), as listed below

(see also Table 4.1):

● trapped electrons—outside spacecraft and inside EVA suits● reentrant and splash albedo electrons—outside and inside

spacecraft, and inside EVA suits● trapped protons (�10 MeV)—do not penetrate spacecraft or

EVA suits● protons and light nuclear particles (�10 MeV)—outside and

inside spacecraft, and inside EVA suits● GCR ions and secondary charged fragments—outside and inside

spacecraft, and inside EVA suits● charged-particle fragments—inside spacecraft and inside EVA

suits● neutrons—outside and inside spacecraft, and inside EVA suits


The relative contributions from each component (including the

secondary radiation) at each location will vary according to several

factors, including the mass distribution inside the spacecraft, the

EVA suit design and materials, and the site of interest within the

human body.

1.7 Approach to Dose Assessment for Astronauts

Recommendation 6: Dose assessment for astronauts should

utilize a combination of radiation transport calculations and

measurements as illustrated in Figure 5.1. The main features

of the approach should include sequential assessment of the

radiation environment at the exterior surface of the space-

craft, the interior radiation environments in the spacecraft

and EVA suits, and the transmission of radiation to internal

organs or tissues in order to estimate the dose-limit quanti-

ties. The radiation transport calculations are not intended

to be a substitute for measured data, but are designed to

augment the measurements such that the combination of

measurements and calculations should provide an estimate

of the dose-limit quantities in Table 2.1 to within a factor of

1.5 at the 95 percent confidence level.

Environmental models for GCR (and the associated albedo neu-

trons), trapped radiations, and SPEs are used to represent the exte-

rior radiation field in LEO. External measurements can be made

outside the spacecraft to allow correction to the models to reduce

uncertainties.

Shielding models for the Space Shuttle, ISS, and EVA suits allow

evaluation of the radiation environment to which the astronaut is

exposed. Except for the absolute intensity of the trapped radiation

or a SPE, the interior radiation environment can be evaluated with

high-speed computational models. The interior radiation environ-

ments of the Space Shuttle and ISS can be monitored with various

instruments and the measurements can be used to adjust the esti-

mate of the trapped-particle intensity, reduce the uncertainty in

the model estimates, evaluate transmission factors, and evaluate

calculated dosimetric quantities. Personal dosimeters can provide

estimates of absorbed dose at points on the surface of the astro-

naut’s body.

The evaluation of organ or tissue doses for astronauts can be

performed with computerized male and female anthropomorphic

models. The models allow the evaluation of the relationships between

1.8 OPERATIONAL RADIATION MONITORING / 7

absorbed dose (D) and dose equivalent (H ) at points on the surface

of the body and the required quantities in deeper-lying organs [i.e.,

mean absorbed dose in an organ or tissue (DT) and organ dose equiva-

lent (HT), the surrogate for equivalent dose (HT)] that are needed to

obtain the dose-limit quantities effective dose (E) and gray equiva-

lent (GT).

In addition, occupancy factors keyed to individual astronaut activ-

ity can be used to estimate exposures from the personal dosimeter

measurements. This is especially true during EVA where large fluc-

tuations in the trapped electron environment or SPEs could occur.

1.8 Operational Radiation Monitoring

Recommendation 7: Operational radiation monitoring con-

sisting of area monitors and personal dosimeters should pro-

vide measured data of sufficient accuracy:

● for determination of field quantities and organ or tis-

sue doses to be used for normalizing radiation transport

calculations

● for dose assessment and record keeping purposes

● for real-time or near real-time estimates of dose rates for

purposes of immediate dose management or ALARA

1.8.1 Area Monitoring

Recommendation 8: Tissue equivalent proportional counters

(TEPC) should be utilized during manned space flight for

real-time measurements of absorbed dose and absorbed dose

rate, and estimates of quality factor and dose equivalent to

a small mass of tissue.

A TEPC is an active detector that is designed to measure energy

deposition in volumes of tissue comparable to the dimensions of the

nuclei of mammalian cells. Data are recorded on an event-by-event

basis such that one can obtain a distribution of biologically relevant

energy deposition events. Its tissue equivalence and large dynamic

range make it sensitive to photons, neutrons and charged particles

from electrons and protons to heavy ions.

When the data are integrated over the complete distribution of

lineal energy (y), TEPCs can generate absorbed dose (D) and ab-

sorbed dose rate (D·). The distribution of energy deposition events

depends on characteristics of the radiation field and the response of


the detector, and can serve as a test for radiation transport models

or to obtain an approximation to the quality factor (Q) for protons

and heavier particles. This detector provides a reliable estimate of

D from protons through high atomic number, high-energy (HZE)

particles as well as photons, electrons and neutrons. Although the

data in terms of y are not a direct replicate of the distributions of

fluence [&(L)] or absorbed dose [D(L)] as a function of LET (L),

average values of y, in particular dose-averaged y, are numerically

similar to L. Data from a TEPC can be displayed continuously and

stored for later transmission to the Mission Control Center.

Recommendation 9: Solid-state detectors should be utilized

during manned space flight for real-time measurements of

LET distributions both inside and outside of the spacecraft.

Solid-state detectors record the energy deposited by a charged

particle. The ratio of the deposited energy to the thickness of the

detector yields the approximate LET for the incident particle. Thus

a single detector can provide data that yield absorbed dose (D) and

absorbed dose rate (D·) for protons and heavier charged particles. It

can be fabricated into a compact detector for use as portable area

monitor or personal dose-rate meter with on-demand readout. Sev-

eral of these detectors can be combined to point in different directions

to provide a more complete description of the radiation field either

outside or inside of the spacecraft.

Recommendation 10: An active detector sensitive to electrons

should be installed outside of the spacecraft to serve as a

monitor for fluctuations in the electron component of the

space radiation environment, which can change by many

orders of magnitude during and following an SPE due to

short-term perturbations of the geomagnetic field.

The fluctuations of the electron component could be of concern

during an EVA, since electrons above a few hundred kiloelectron

volts can penetrate the space suits. Such a monitor could be a simple

ionization chamber with a wall thickness sufficient to attenuate very

low-energy electrons but thin enough to record electrons that could

penetrate a space suit.

1.8.2 Personal Dosimetry

Recommendation 11: A measurement package consisting of a

thermoluminescent dosimeter (TLD) or optically stimulated

1.8 OPERATIONAL RADIATION MONITORING / 9

luminescence dosimeter (OSLD) for measurement of the low-

LET component, and a stack of plastic nuclear track detec-

tors (PNTD) to determine the high-LET component should

be used for passive personal dosimetry in the complex radia-

tion field experienced in space.

LiF:Mg,Cu,P (lithium fluoride, doped with magnesium, copper and

phosphorus) would appear to be an attractive TLD material. Alterna-

tively, Al2O3:C (aluminum oxide, doped with carbon) is the best cur-

rently available OSLD material. To measure the dose equivalent (H)

from the high-LET components, polyallyl diglycol carbonate [PADC/

CR-39� (trade name CR-39, PPG Industries, Inc., Pittsburgh, Penn-

sylvania)] is the PNTD material proposed. Such devices have been

used as part of the area monitoring or personal dosimeter packages

on the Space Shuttle, but are not currently planned for ISS. It is

recommended that they be used as personal dosimeters on both

vehicles. CaF2:Tm (calcium fluoride, doped with thulium) [e.g., TLD-

300� (Bicron, Saint-Gobain Industries, Cleveland, Ohio)] could also

be used as an adjunct personal dosimeter to provide additional infor-

mation to normalize radiation transport models, but not for quantita-

tive determination of dose quantities.

With these detection elements in a passive personal dosimeter

package, H at a point in adjacent tissue is then obtained by using

a combination of TLDs (or OSLDs) and PNTDs, as described in

Section 6.3. In this recommendation, TLDs (or OSLDs) are used to

measure D in the low-LET region (L � 10 keV �m�1) for which

Q � 1. It is further recommended that D in the high-LET region

(L � 10 keV �m�1) be monitored using PNTDs. In this region Q

is dependent on L. Correction may be needed for any overlap of

the two responses so that intermediate LET components are not

double-counted.

Verification of LET-dependence of the TLD response of LiF:

Mg,Cu,P and of the OSLD response of Al2O3:C will be required. Until

such time as these data are available, LiF:Mg,Ti-based dosimeters

[e.g., TLD-100� or TLD-700� (Bicron, Saint-Gobain Industries,

Cleveland, Ohio)] may still be used, along with PNTDs, in order to

provide LET data suitable for correcting the TLD dose response for

the L � 10 keV �m�1 component, and for estimating H for this

component from the PNTD results. If PNTDs cannot be used, a

different, less desirable approach has to be adopted using TLDs

and data from a TEPC or particle spectrometer, as described in

Section 6.3.4.2.

For purposes of active personal dosimetry, a thick silicon detector

may be able to provide an approximation to D, D�, and D(L) for


protons and heavier particles. These have been designed in a suffi-

ciently compact configuration to be worn by the astronauts and can

be read on demand. In the case of currently available active personal

electronic dosimeters, most are used routinely to measure low-LET

radiation and have not been characterized for the types and energies

of particles comprising the fields in spacecraft. Such dosimeters,

when well characterized, may perform a useful role. Future consider-

ations for active electronic personal dosimeters are noted in Sec-

tion 6.3.2.3.

In those cases where active personal dosimeters are not used,

onboard systems for analysis of passive personal dosimeters may be

required, especially on long-duration ISS flights. Onboard systems

for readout of TLDs and OSLDs are certainly possible. However,

onboard readout of PNTDs is not feasible. Therefore, onboard read-

out of passive dosimeters will provide only part of the dose record

(for low-LET) and development of such systems should only be con-

sidered if active personal dosimetry is unavailable.

The potential for developing a set of conversion coefficients that

directly relate H obtained with TLDs and PNTDs at the surface to

E for the space radiation environment, similar in concept to those

used in other occupational radiation environments, would be worth

investigating.

1.8.3 Calibration

Recommendation 12: Response data for the active and passive

devices used should be determined for the following energy

ranges as appropriate: protons from 10 to 800 MeV; high-Z,

high-energy ions (e.g., helium, carbon, silicon, iron) from

50 MeV n�1 to 1 GeV n�1; electrons from 0.5 to 10 MeV; and

neutrons from 1 to 180 MeV, in fields that are monoenergetic

or quasi-monoenergetic, plus response data for fields which

replicate the neutron field produced by the interactions of

GCR with shielding material.

The response characteristics of all the types of devices should be

determined prior to use. This will normally be accomplished by a

combination of experiment and calculation. The response determina-

tions should normally be in terms of the quantity fluence. An excep-

tion would be for the determination of photon response, for which

air or tissue kerma will be more appropriate. For the determination

of the response characteristics of personal dosimeters, some irradia-

tions should be performed on either an anthropomorphic phantom

or a surrogate. Sufficient angle dependence of response data should

1.10 IMMEDIATE DOSE MANAGEMENT / 11

be available to estimate the isotropic response. Where needed and

where available, recommended fluence to D and fluence to H conver-

sion coefficients should be used.

1.9 Biodosimetry

Recommendation 13: NASA should continue to use biodosime-

try as an ancillary component of radiation dose assessment

for astronauts during extended space flights.

The unique contribution of a biodosimetry program is that it pro-

vides an individual’s dose assessment as estimated from a biological

endpoint. Thus, it includes the response to the cumulative exposure

and allows for an assessment of variations in individual sensitivity.

The fluorescence in situ hybridization (FISH) method is the most

appropriate approach based upon available knowledge, technical

availability, and experience. The current approach of using FISH

for analyzing stable chromosomal translocations in peripheral lym-

phocytes both in preflight and postflight samples appears to be

providing useful information on exposures. The establishment of

calibration curves from individual preflight blood samples increases

the sensitivity. Incorporating analysis of prematurely condensed

chromosomes (PCC) will provide more analyzable cells within a sam-

ple. Improvements that can be envisaged are using chromosome

painting probes and computer analysis that allow for assessment of

translocations in all chromosomes at the same time. This method

has been used successfully for tumor analysis. Automating the vari-

ous FISH methods will increase throughput enormously.

Future considerations for biodosimetry in the area of genomics

or molecular profiling, and technologies for measuring changes in

cellular markers (in response to radiation) are noted in Section 7.

1.10 Immediate Dose Management and ‘‘As Low As

Reasonably Achievable’’

Immediate dose management refers to actions taken to address

high transient exposures in the space radiation environment that

could impact the conduct or completion of the mission or mission

tasks.

Recommendation 14: Implementation of immediate dose man-

agement actions is the responsibility of all team members


involved with work impacting the astronaut’s exposure to

radiation. A written plan (notably the flight rules mecha-

nism) should contain the implementing procedures.

ALARA refers to actions taken to keep the doses in all cases as

low as reasonably achievable, by balancing the mission objectives

with practical dose reduction steps.

Recommendation 15: The RHO should assess the opportuni-

ties to apply ALARA. However, an effective ALARA program

depends on everyone involved in the design and management

of spacecraft and missions understanding the space radia-

tion environment and its impact on astronaut radiation expo-

sure. ALARA concepts should be incorporated into the design

of the spacecraft and suits, the preflight planning (including

the planned in-flight procedures), an in-flight review, and a

postflight review.

A number of suggestions bearing on immediate dose management

and ALARA are given in Section 8.4. Three examples are:

● place radiation instruments at locations that provide the best

real-time information on radiation exposure to the astronauts

(for both immediate dose management and ALARA);

● provide areas where astronauts could be moved during high

transient exposures, i.e., move to a safe haven with additional

shielding, and/or reposition the Space Shuttle (for immediate

dose management); and

● provide areas used during off-duty hours and sleeping quarters

with optimized shielding (for ALARA).

1.11 Radiation Safety Training

Recommendation 16: All personnel whose work impacts on

astronaut radiation exposure should be trained in the tech-

niques of radiation protection, with emphasis on implemen-

tation of immediate dose management and ALARA.

The scope and depth of this training should be related to the

corresponding level of impact the individual may have on astronaut

dose. Section 8.5 presents suggested radiation protection training

topics for astronauts, flight directors, flight surgeons, RHOs, other

radiation safety professionals, and other individuals whose work can

affect the astronaut’s radiation exposure. All of these individuals

should be trained in NASA’s operational radiation safety program

1.12 DOSIMETRY RECORD / 13

and how immediate dose management and ALARA actions are

proposed, implemented and made part of the review of actual events,

especially those events involving high transient exposures.

Recommendation 17: Astronauts should be trained in the

proper wearing and care of personal dosimeters. NASA

should have a clear requirement and related training for the

use of personal dosimeters by astronauts in-flight to ensure

that each astronaut has an accurate mission and career

dose record.

1.12 Dosimetry Record

Recommendation 18: The dosimetry record constitutes the

formal documentation for each astronaut’s space-related

radiation exposure history and should contain the cumula-

tive dose from space flight, mission-related aviation activi-

ties, and mission-related biomedical research. The dosimetry

record should contain, or be linked to, all the basic informa-

tion that is necessary to obtain the required dose-limit quan-

tities [gray equivalent (GT) and effective dose (E)], and should

include the low- and high-LET components of the radiation

field.

The dosimetry record, and the other supporting records linked to

it, should be kept in a manner to satisfy a number of purposes

as described in Section 9. These records are important to protect

astronauts and to document radiation exposures. Other radiation

exposure files, such as diagnostic and therapeutic medical radiation

exposures from overall health care that are maintained in the medi-

cal department, should also be linked to the dosimetry record. How-

ever, the diagnostic and therapeutic medical radiation doses should

not be added to occupational doses either for planning purposes or

to limit occupational exposures.

Recommendation 19: Astronauts should receive an annual

confidential report on their radiation dose assessment. The

report should include career radiation doses in terms of effec-

tive dose (E) for stochastic effects and monthly and annual

doses in terms of gray equivalent (GT) for deterministic

effects.

Recommendation 20: The dosimetry record should be

updated retrospectively whenever there is a systematic


change in methodology or new information becomes avail-

able. Astronaut dose estimates should be adjusted whenever

differences in the revised dose assessments exceed 30 percent

of the original dose assessment (see discussion of accuracy

in Sections 6.4.1 and 6.4.2).

Existing data related to space missions should be examined and

compared to the dosimetry record to confirm that radiation dose

estimates are based on all available data. If previously unanalyzed

data are found, the data should be identified and representative

samples of the data should be fully analyzed to determine the extent

of their effect on current dose estimates.

2. Objectives of theOperational RadiationSafety Program forAstronauts

There are currently 160 astronauts in the U.S. Astronaut Office

(137 United States astronauts and 23 other international astronauts)

and approximately 40 Russian cosmonauts. The number of United

States astronauts involved in the foreseeable future in LEO activities

is anticipated to be between 200 and 250. Therefore, the operational

radiation safety program for astronauts is for a small population.

The overall objective to assess and control the radiation exposure

of individual astronauts can be broken down as follows: (1) keep

individual doses below the established dose limits to avoid determin-

istic effects; (2) keep accumulated doses over an astronaut’s career

below the established dose limits for stochastic effects; and (3) keep

all astronaut doses as low as reasonably achievable, economic and

social factors being taken into account (i.e., follow the standard

ALARA principle of radiation protection). In the context of near-

Earth space activities, one must also take into account the mission

requirements and the prevailing radiation conditions in LEO.

2.1 The Low-Earth Orbit Program

There are unique considerations in balancing mission objectives

in LEO against the resulting levels of radiation exposure. The deci-

sion to incorporate Russian launch capabilities into the construction

and operation of ISS placed the Station in a high-inclination orbit,

comparable to the previous Russian Mir Space Station. The higher

inclination places ISS in orbits that increase the residence time in

radiation zones where the probability for high-exposure events from

solar energetic particles and penetrating electrons in the outer

radiation belts to occur is higher, compared to the original plan

for a low-inclination orbit, which would have mostly avoided these

radiation zones (NAS/NRC, 2000). This is a particularly important

15

16 / 2. OBJECTIVES OF THE SAFETY PROGRAM

consideration given the extensive extravehicular activities planned

during ISS construction, which will also span the maximum of the

current 11 y solar cycle.

A helpful description of the impact of the high-inclination orbit in

which the astronauts will work is given in a report by the National

Academy of Sciences/National Research Council (NAS/NRC, 2000).

In the quoted text below, abbreviations have been spelled out in

italics.

‘‘As originally conceived in the early 1980s, ISS [the International

Space Station] was to have a low-inclination (28.5 degrees), low-

altitude (350 km) orbit. Then, the SAA [South Atlantic Anomaly . . .

part of the Van Allen Belts] and GCRs [galactic cosmic radiations]

would have been the only significant sources of radiation. SRAG [the

Space Radiation Analysis Group . . . the operational radiation safety

unit at Lyndon B. Johnson Space Center] knows how to design mis-

sion schedules to minimize astronaut exposure to the SAA during

EVAs [extravehicular activities . . . astronauts in space suits outside

a Space Shuttle or station], and there is little anyone can do to

minimize GCR exposure. In 1993, however, the United States agreed

with the Russian Federation to incorporate Russian launch capabili-

ties into ISS construction and maintenance. That agreement brought

with it the need to place ISS in a high-inclination orbit, essentially

the same as that of the Mir space station, 51.6 degrees geographic

[350 to 450 km altitude]. Consequently, ISS and the astronauts who

construct and use it run the risk of being exposed to solar energetic

particles and penetrating electrons in the horns of the outer belt.

Exposure from these sources will be sporadic since SPEs [solar parti-

cle events] follow solar storms and HREs [highly relativistic electrons]

follow magnetic storms and impacts by strong solar wind shocks.

During the declining phase of a solar cycle—perhaps late in ISS

construction—HRE events are also associated with times, lasting

about a week, when solar wind streams are especially fast. Whereas

satellite encounters with the SAA are as predictable as the tides,

usually solar energetic particles, geomagnetic storms, and high-

speed solar wind streams are not reliably predictable, nor is the

intensity of the associated radiation event. The high-inclination orbit

of ISS therefore introduces a new radiation risk factor.

‘‘ISS construction plans call for approximately 33 U.S. shuttle

flights and 10 Russian flights. The construction phase will extend

from 1998 to 2004, which spans the maximum of solar cycle 23, when

SPEs are expected to be most frequent. NASA estimates that during

those years astronaut and cosmonaut construction crews may have

to perform more than 160 EVAs totaling more than 1,100 hours.

During those same years, there will be more than 400 additional


hours of EVAs by astronauts and cosmonauts to service and main-

tain the station. The total exceeds 1,500 hours, or 1,000 ISS orbits,

of EVA time.’’

The very real potential that high transient radiation doses to

astronauts will occur occasionally during construction and operation

of ISS was also documented by NAS/NRC (2000). The following two

statements from the report address the potential for high radiation

doses from SPEs and relativistic electrons from the outer belt,

respectively.

‘‘. . . the high-latitude zones to which solar energetic particles have

access show a marked tendency to widen over the polar latitudes

reached by the ISS orbit when SPEs are in progress, a tendency that

becomes more pronounced as SPEs intensify. Two storms during

1989, near the maximum of the last solar cycle, illustrate the point.

The areas around the poles accessible to SPE particles enlarged until

they engulfed more than a quarter of the ISS orbit, and the flux of

particles was high enough to have pushed an astronaut over the

short-term limit for irradiation of skin and eyes during a single ill-

timed 6-hour EVA.’’

‘‘For a portion of nearly every day, some fraction of the ISS orbit lies

within the outer radiation belt, where relativistic electrons reside. At

its maximum, this fraction is about 20 percent. During occasions

called relativistic electron events, which happen on average about

once per month and last several days, the intensity of relativistic

electrons in the belt increases by up to four orders of magnitude.

When the intensity of relativistic electrons is greatest, a single ill-

timed EVA could deliver a radiation dose big enough to push an

astronaut over the short-term limit for skin and eyes.’’

NAS/NRC (2000) also made recommendations for improving the

real-time forecasting of space weather events such as solar wind

conditions (particularly during the current 11 y solar maximum

cycle) and outer radiation belt conditions, and the need for improved

coordination of existing and future space weather forecasting capa-

bilities within NASA, and with the National Oceanic and Atmo-

spheric Administration and the U.S. Air Force. Most of these

improvements will take some time to implement.

2.2 Dose Limits for Astronauts

NCRP reviewed the evolution of dose limits for astronauts and

recommended dose limits specifically for astronauts in LEO in NCRP

Report No. 98 (NCRP, 1989). NCRP recently updated the dose limits


in NCRP Report No. 132 (NCRP, 2000a) to take account of more

recent information on estimates of radiation health effects. The dose

limits for bone marrow, lens of the eye, and skin are for protection

against deterministic effects. The career dose limits are for protection

against delayed stochastic effects and are based on a lifetime excess

risk of cancer mortality of three percent.1 The main features of the

current dose limits and the quantities used to assess astronaut expo-

sures against the dose limits are summarized in Table 2.1.

2.2.1 Deterministic Limits

The dose limits for deterministic effects are expressed in terms of

gray equivalent, where gray equivalent is the mean absorbed dose

in an organ or tissue modified by a recommended value, for radiation

protection purposes, of the relative biological effectiveness (i.e., a

best estimate) of a given particle type, as given in NCRP (2000a).2

In this Report, a conventional notation is used for the quantity gray

equivalent, namely:

GT � RiDT, (2.1)

where GT is gray equivalent, Ri is the recommended value of the

relative biological effectiveness for particle type i referred to above,

and DT is the mean absorbed dose in an organ or tissue.

The values for Ri are given in Table 2.2. The dose limits for deter-

ministic effects are given in Table 2.3. The special name for the unit

of DT is gray (Gy) and the name of the unit for GT is gray equivalent,

with the notation Gy-Eq (NCRP, 2000a).

2.2.2 Stochastic Limits

The career limits for delayed stochastic effects are expressed in

effective dose (E), where:

E � �T

wTHT. (2.2)

1The recommended dose limit for an astronaut entails a similar lifetime excess risk

of cancer mortality (three percent) as the dose limit for a terrestrial worker, if both

an astronaut and a terrestrial worker reached the maximum value of the respective

dose limit during their working lifetimes.2NCRP (2000a) recommended that the mean absorbed dose in an organ or tissue

(DT) be modified by the recommended values for relative biological effectiveness to

adjust for radiation quality for deterministic effects because the usual formulation

for equivalent dose (HT) is obtained by applying radiation weighting factors (wR) which

are applicable to stochastic effects.


TABLE 2.1—Radiation protection quantities for space radiation

(NCRP, 2000a).

Quantity Deterministic Effects Stochastic Effects

Mean absorbed Mean absorbed dose Mean absorbed dose (DT)

dose in an organ (DT) (in Gy) (in Gy)

or tissue For irradiated For entire organ or tissue

portion of tissue

DT weighted for Gray equivalent (GT) Equivalent dose (HT)

the relative (in Gy-Eq) (in Sv)

biological Weights DT by a For space radiation,

effectiveness of recommended value adopts the relationship

the radiation or for relative biological HT � HT (NCRP, 2000a),

particle type effectiveness for the where HT is the organ

particle type i (Ri); dose equivalent (ICRU,

Ri values derived 1993) (Section 2.2.2).

from ICRP (1989). Uses the Q(L)

Tissues are lens of relationship in ICRP

eye, skin and bone (1991), ICRU (1993), and

marrow (Table 2.2) NCRP (1993) to evaluate

HT (Section 2.2)

Tissue weighting Not applicable Uses the tissue weighting

factor factors for radiation

detriment (wT) from ICRP

(1991) and NCRP (1993)

Dose-limit Gray equivalent (GT) Effective dose (E) (in Sv)

quantities (in Gy-Eq) E � �T

wTHT, from ICRP

(1991) and NCRP (1993).

HT is the approximation

for HT, determined as

given above

Dose limits GT limits (in Gy-Eq) E limits (in Sv)

For 30 d and annual, For 10 y career, gender

all three tissues. and age-specific, males

For career, lens of and females, for ages at

eye and skin (Table exposure of 25, 35, 45

2.3) and 55 y.

Based on three percent

excess lifetime risk of

cancer mortality (Table

2.4)


TABLE 2.2—Ri values for converting DT to GT for deterministic

effects (adapted from ICRP, 1989; NCRP, 2000a).a

Particle Type Ri (range for value)

1 to 5 MeV neutrons 6.0b (4–8)

5 to 50 MeV neutrons 3.5b (2–5)

Heavy ions (e.g., helium, carbon, neon, argon) 2.5 c (1–4)

Protons �2 MeV 1.5 (—)

aRBE values3 for late deterministic effects are higher than for early effects

in some tissues and are influenced by the doses used to determine the

RBE values.bThere are not sufficient data on which to base an Ri for early or late

effects4 induced by neutrons of energies �1 MeV or greater than about

25 MeV. However, based on the induction of chromosome aberrations, using

250 kVp x rays as the reference radiation, RBE values for neutrons �1 MeV

are comparable to those for fission spectrum neutrons. It is reasonable to

assume that RBE values for neutrons �50 MeV will be equal or less than

those for neutrons in the 5 to 50 MeV range.cThere are few data for the tissue effects of ions with Z � 18, but RBE

values for iron ions (Z � 26) are comparable to those for argon. Based on

the available data, an Ri of 2.5 for heavy ions is reasonable. One possible

exception is cataract of the lens of the eye because high RBE values for

cataracts in mice have been reported.

HT is the equivalent dose and wT is the tissue weighting factor (ICRP,

1991; NCRP, 1993). The career limits are specified by gender and

age, and take into account the gender and age-adjusted cancer mor-

tality risk coefficients (i.e., the differential risk per unit dose) for the

same overall three percent excess lifetime risk. The career dose limits

are given in Table 2.4.

The quantity E is obtained as follows for the space radiations.

First, the dose equivalent (H ) is defined at a point in tissue and can

be directly measured (ICRU, 1993). H is given by:

H � �L

Q(L) D(L) dL. (2.3)

Q(L) is the quality factor (Q) for particles as a function of LET (L)

and D(L) is the spectral distribution in terms of L of the absorbed

3The acronym RBE refers to experimental data on relative biological effectiveness.4The symbol Ri refers to recommended values of relative biological effectiveness

(i.e., best estimates) for radiation protection purposes that are used for obtaining GT,

as discussed in NCRP (2000a).


TABLE 2.3—Recommended GT limits for deterministic effects

(all ages) (NCRP, 2000a).

GT Limit

(Gy-Eq)

Bone Marrow Lens of the Eye Skin

Career —a 4.0 6.0

1 y 0.50 2.0 3.0

30 d 0.25 1.0 1.5

aThe career limits for stochastic effects given in Table 2.4 are considered

to be more than adequate for protection of bone marrow against deterministic

effects for a career. The career limits are expressed in terms of E. The wR

values (for stochastic effects) used to convert DT to E are based on Q(L) and

are higher than the Ri values (for deterministic effects) used to convert DT

to GT. Therefore, there is no need for a career deterministic limit for the

bone marrow. The career stochastic limit is more restrictive and would

always be expected to result in a lower DT to the bone marrow for irradiation

conditions in space.

TABLE 2.4—Ten-year career E limits based on three percent excess

lifetime risk of fatal cancer (NCRP, 2000a).a,b

E Limit

(Sv)Age at Exposureb

(y) Female Male

25 0.4 0.7

35 0.6 1.0

45 0.9 1.5

55 1.7 3.0

aA three percent excess lifetime risk of cancer mortality has additional

components of detriment associated with it, namely the risk of heritable

effects (0.6 percent) and the detriment associated with nonfatal cancer (also

0.6 percent), for a total detriment of 4.2 percent. These nominal risks are

as given in ICRP (1991) and NCRP (1993).bThe career E limits are for a 10 y career beginning at the age of first

exposure. For situations in which careers differ in length from 10 y, or for

careers that start at other than the designated ages, additional advice is

provided in Section 6.4 of NCRP (2000a).


dose at the point (D). For the complex mixtures of high- and low-LET

radiations experienced in LEO, the practice in the space radiation

protection community is to average the point quantity H over the

organ or tissue of interest by means of computational models to

obtain the organ dose equivalent (ICRU, 1993), which is required

for radiation protection purposes but cannot be directly measured.

In this Report, the symbol HT is used for the quantity organ dose

equivalent. NCRP (2000a) adopted HT as an acceptable approxima-

tion for HT for stochastic effects: Therefore, in general terms:

HT � MT�1 �

x�

L

Q(L) D(L) �(x) dL dx, (2.4)

where there is a second integration over the points x in tissue T with

tissue density �(x) and total mass MT. For the radiations in LEO,

the quantities HT and HT are interchangeable for radiation protection

purposes. The methods used to evaluate HT are discussed in Sections

5 and 6. Therefore, E � wT HT, and E, H, HT and HT are expressed

in sievert (Sv).5

The Q(L) relationship is given in ICRP (1991), ICRU (1993), and

NCRP (1993), where:

Q(L) � 1 for: L � 10 keV �m�1

Q(L) � 0.32 L � 2.2 for: L � 10 to 100 keV �m�1 (2.5)

Q(L) � 300 L�1/2 for: L � 100 keV �m�1

The Q(L) relationship is consistent with current knowledge of the

general trend of relative biological effectiveness for cell killing and

induction of mutation from HZE particles.

2.3 Operational Radiation Protection Considerations

This Report concentrates on the specific technical considerations

necessary to implement an operational radiation safety program for

the principles of dose limitation and ALARA for astronauts working

in LEO. In particular, the Report describes the radiation components

5This procedure for obtaining E for space radiations differs from that used for

terrestrial radiation environments, in that HT (calculated as given above) replaces

HT � wR DT, where wR (for external radiation) is a nominal value based on the type

and energy of the radiation incident on the body (ICRP, 1991; NCRP, 1993). The

practice of obtaining HT permits more complete consideration of the Q(L) relationship

for the complex space radiation environments.

2.3 OPERATIONAL RADIATION PROTECTION CONSIDERATIONS / 23

of the space environment (Section 4) from the perspective of the

measurements needed to support the dose assessment approach (Sec-

tion 5) used to generate values for the radiation protection quantities

recommended for space activities, that is, GT, HT (i.e., the surrogate

for HT), and E. A key aspect of the Report is the recommendations

for collection of observable data (Section 6) in both active and passive

modes. The observable data help implement the recommended dose

assessment approach, in conjunction with an array of radiation

transport models and codes (Section 5). These recommendations are

developed fully in this Report, taking into account the practical

limitations on the radiation detection devices and protective systems

that can be available in-flight.

The availability of credible estimates for the recommended radia-

tion protection quantities would help accomplish the following objec-

tives of the operational radiation safety program:



that in turn significantly influence the levels of radiation expo-

sure to the astronauts, and radiation protection decisions that

in turn significantly influence the conduct of the mission;

● to collect and record astronaut doses for individual mission and

cumulative career records. These records can also be used in

prospective radiation-related health studies of the Astronaut

Corps or to assist retrospective dosimetry to update the records

for previous activities of individual astronauts; and

● to identify, plan and inform practical ALARA actions to avoid

unnecessary levels of radiation exposure. These ALARA actions

would address both radiation exposures that are adjunct to a

mission (e.g., ground-based space-related biomedical research)

and those that occur in-flight (e.g., mission experiments using

radiation sources, or avoidance of higher exposure locations in

space vehicles when an ongoing mission task or activity does

not require presence there).

3. Current Management ofAstronaut RadiationSafety Program

The management of radiation dose received by astronauts in LEO

involves many individuals. Typically, astronauts do not play an

active role in decision making and policy regarding radiation issues.

Instead, the flight director and flight surgeon direct their actions,

with the help of radiation experts.

At the Lyndon B. Johnson Space Center (JSC), there are two

radiation expert resources, SRAG and the RHO. These individuals

work with the medical officers, the payload officers, mission planners,

outside agencies, and the flight directors to provide radiation exper-

tise in developing and interpreting radiation flight rules that govern

the conduct of a flight, from mission planning through the end of a

space flight, with regard to crew radiation exposure and adherence

to ALARA. Dose limits and administrative levels are developed and

are then written as flight rules by NASA radiation experts.6 The

flight rule approval process involves a variety of NASA managers

from different departments so that management becomes very famil-

iar with the rules and supporting rationale. In the process of approv-

ing flight rules, program management is educated and trained on

the risks associated with radiation exposure. A flight director chairs

the Flight Rule Change Board, which approves and incorporates all

new rules and rule changes. Once approved and incorporated, the

flight rules are ‘‘tested’’ during simulations to ensure proper interpre-

tation and response to a given scenario.

3.1 Flight Rules for Management of Dose

The flight rules are a comprehensive set of rules, reviewed and

updated continuously, governing all the procedures and operation

6 Dose limits are developed through input from NCRP and the Occupational Safety

and Health Administration recommendations. Administrative levels are developed

internally by NASA.

24

3.1 FLIGHT RULES FOR MANAGEMENT OF DOSE / 25

of space vehicles. Extensive rules regulate management of normal

operations and off-nominal operations of all Space Shuttle and ISS

systems, such as main engines, life support, computer, electrical,

and hydraulic systems. The flight rules that govern measurement

of the radiation environment and management of crew exposure to

radiation are contained in one section of the flight rules titled ‘‘Space

Environment,’’ which incorporates the following:

● general definitions which define event conditions (e.g., SPE,

energetic SPE, geomagnetic storm) and ALARA;

● radiation subsystem loss definitions which define what deter-

mines monitoring and measuring equipment to be out of service;

● crew exposure management which defines actions to assist in

management of the crew exposures;

● rules that are in place to maintain exposure ALARA and below

legal limits;

● radiation subsystem management which are special rules for

radiation equipment operation; and

● designated maintenance items which are additional criteria that

indicate hardware may be inoperable.

Administrative limits or ‘‘action levels’’ are published in the flight

rules to serve as a guide, with ALARA, to initiate more stringentdose management activities for a mission if there is a higher thanprojected accumulated exposure. The administrative limits for a mis-sion are monthly, cumulative limits 5 mGy above the projectedabsorbed dose for that part of the solar cycle. There are also 30 dand 1 y maximum allowable exposure limits to manage acute expo-sures. If the exposure on a particular mission is greater thanexpected, actions are detailed in the flight rules to remain withinthe crew administrative limits. Some actions that the ground andcrew might consider are: restricting crew location within ISS,rescheduling EVAs, reducing the number of EVAs, terminatingan EVA in progress, deferring ISS reboost, shortening crew timein orbit, or returning the crew to Earth. Crew safety is the high-est priority. Many factors determine the course of action during aflight. SRAG, the RHO, and the flight surgeon work with the flightdirector to determine the best course of action when these rules areconsidered.

The NASA term ‘‘administrative limit’’ or ‘‘action level’’ corres-ponds to the term ‘‘administrative level’’ used elsewhere in thisReport. The NASA term ‘‘exposure limit’’ corresponds to the term‘‘dose limit’’ used elsewhere in this Report.

There is a set of flight rules governing the conduct of a SpaceShuttle mission (e.g., NASA, 2000a), and a set governing the conductof an ISS mission (e.g., NASA, 2000b).

26 / 3. CURRENT MANAGEMENT OF SAFETY PROGRAM

3.2 Biomedical Research

Because astronauts may also receive occupational radiation dose

during the conduct of biomedical experiments associated with space

flights, these experiments need to be approved for funding and perfor-

mance during flight at NASA Headquarters. Exposure to radionu-

clides and x rays associated with biomedical experiments may

contribute substantially to an astronaut’s total dose. Any biomedical

experiment involving radiation exposure is approved by an expert

NASA radiation panel. The RHO, a flight surgeon, and a senior

medical officer are among those who serve on this panel. After

approval by this group, the JSC Institutional Review Board reviews

the experiment and may give approval. As with all biomedical experi-

ments, astronauts may choose whether or not to volunteer to partici-

pate. If astronauts choose to volunteer, they do so with informed

consent.

3.3 Individuals Involved in Management of Dose

3.3.1 Astronaut

Approximately 160 astronauts are active in the U.S. Astronaut

Office. One or two astronauts (generally astronauts with medical

backgrounds) represent the Astronaut Office’s position regarding

radiation protection policy, procedures and concerns in the various

meetings and committees. Prior to their flight, if radiation exposure

is expected to be more than minimal, the affected astronauts may

participate more actively in the decision-making process concerning

their particular flight. In general, astronauts assigned to fly longer

missions on ISS show more interest in radiation protection issues

and take a more active role in developing procedures to reduce expo-

sure, such as determining the ‘‘best’’ sleep location.

3.3.2 Flight Director

Approximately 20 flight directors are stationed at JSC. Each mis-

sion is assigned a ‘‘lead’’ flight director who provides input and makes

decisions during the planning and training period preflight. During

a mission, the flight director is the final decision maker in the Mission

Control Center with regard to all aspects of the mission including

mission replanning, and flight rule interpretation and implementation.

3.3 INDIVIDUALS INVOLVED IN MANAGEMENT OF DOSE / 27

The flight director weighs all risks and benefits during this process

with the aid of systems experts on the flight control team and relies

heavily on the RHO, SRAG and the flight surgeon when decisions

regarding radiation protection issues need to be made. The flight

director weighs the radiation risks against the risks and mission

impacts of altering other flight procedures (for example, delayed or

shortened EVA, or early deorbit and landing at a contingency landing

site). The Chief of the Flight Director’s Office chairs the NASA Flight

Techniques Board that approves flight rules.

3.3.3 Flight Surgeon

Approximately 15 NASA flight surgeons are stationed at JSC.

When serving as a crew surgeon for a specific flight crew, the flight

surgeon is responsible for that crew’s health and safety during all

aspects of flight training, mission execution, and postflight rehabili-

tation and recovery. The flight surgeon briefs the crew preflight

and postflight regarding radiation protection issues and personal

radiation dose. During a space flight the flight surgeon is the primary

interface between SRAG and the flight director in the Mission Con-

trol Center when issues arise pertaining to radiation exposure. Dur-

ing a mission, the flight surgeon and SRAG provide expert flight

rule interpretation for the flight director and mission management,

and assist in mission replanning, when necessary. Generally, one

member of the Flight Surgeon’s Office is designated to be the ‘‘resi-

dent expert’’ on radiation protection issues. This flight surgeon serves

on various committees and groups that address radiation protection

issues. One committee is the Radiation Integrated Product Team,

which is co-chaired by the RHO and the flight surgeon and also

includes experts outside of NASA. This committee meets quarterly

to discuss and formulate possible flight research projects, and to

consider operational issues, such as adequacy of onboard radiation

monitoring. This flight surgeon is also the point of contact to commu-

nicate new information to the rest of the flight surgeon community.

3.3.4 Radiation Health Officer

The RHO provides recommendations to minimize crew exposures

through design and development of radiation protection strategies,

in NASA policy documents, during mission planning, and real-time

during space missions. The RHO tracks crew exposures against

career limits, provides risk interpretation for acute exposures, and

interfaces with the RHOs from the international partners.

28 / 3. CURRENT MANAGEMENT OF SAFETY PROGRAM

3.3.5 Space Radiation Analysis Group

SRAG, along with the RHO, are the NASA radiation specialists

responsible for promoting ALARA and ensuring compliance with

procedures to implement ALARA during all phases of mission design,

planning and flight. They serve as the interface to all other groups

and individuals in the space flight program who have responsibilities

that can affect crew radiation exposure. With the exception of the

flight surgeons, no other individuals involved in mission planning

and operational decision making have in-depth training or experi-

ence in radiation protection issues. During space flights, SRAG has

the following responsibilities: monitoring of space weather; space

radiation environment contingency response and analysis; radiation

instrument operations; tracking of crew daily and cumulative dose;

EVA planning, support and monitoring; and interfacing with outside

support groups (i.e., National Oceanic and Atmospheric Administra-

tion, U.S. Department of Defense). During a mission, SRAG inter-

faces with mission planners and flight controllers, including the

flight director, to update mission significant radiation events and

provide expertise, particularly when transient events in the space

radiation environment produce a potential for high doses to the

astronauts.

4. Radiation Environmentin Low-Earth Orbit

The radiation environment external to a spacecraft in LEO con-

sists of electrons, positrons, neutrons, protons and heavier nuclei

(up to Z � 92). Sources include: GCR from deep space; SPEs produced

by coronal mass ejections or by acceleration in solar flare events;

particles trapped in Earth’s magnetic field; and scattering from

Earth’s atmosphere (albedo neutrons, electrons and protons). Energ-

ies range from a few electron volts for trapped electrons and albedo

neutrons to in excess of 1014 MeV for GCR ions.

The space radiation environment is modulated by spatial and tem-

poral factors including the 11 y solar cycle and the solar wind, and

Earth’s magnetic field, which traps some particles and deflects oth-

ers. The particle distributions are sensitive to both inclination and

altitude. GCR is approximately isotropic in deep space, but in LEO

the GCR intensity is modulated by the geomagnetic field. Due to the

shape of Earth’s magnetic field lines, high-inclination Space Shuttle

flights and ISS (51.6 degrees) are exposed to higher GCR fluxes than

are low-inclination Space Shuttle flights (e.g., 28.5 degrees). Earth’s

magnetic field varies in strength and configuration over timescales

from days to years. It is influenced by changes in activity on the

surface of the sun over the 11 y solar cycle, but can also exhibit acute

changes as a result of solar storms. The January 1997 geomagnetic

storm moved the South Atlantic Anomaly (SAA) by nearly 80 km

(Badhwar, 2001).7

Protons and electrons are trapped in belts at various altitudes

and inclinations. If a spacecraft orbit intersects a belt, the intensity

of the charged particles and the corresponding dose rate onboard

will increase. These belts are slowly modulated by solar activity;

however, the major influence on dose rate is the frequency and dura-

tion of spacecraft transits through the belts, which are on the order of

minutes to tens of minutes in LEO. Active dosimeters can distinguish

between trapped particles and GCR by separating the data into time

7 Badhwar, G.D. (2001). Personal communication (Lyndon B. Johnson Space Center,

National Aeronautics and Space Administration, Houston, Texas).

29

30 / 4. RADIATION ENVIRONMENT IN LOW-EARTH ORBIT

periods corresponding to the parts of the orbit that intersect the

trapped radiation belts.

Coronal mass ejections or flares can generate high intensities of

charged solar energetic particles, primarily protons, that may even-

tually intercept the spacecraft. The occurrence of SPEs is unpredict-

able. Real-time detectors are needed to confirm the onset and

duration of an event in order to initiate countermeasures in the

spacecraft and to avoid high exposures during an EVA.

Although an SPE may last for a short period of time, disturbances

of the trapped belts, and consequently elevated radiation levels, in

particular the intensity of electrons outside of the spacecraft, can

persist for times on the order of several days. Most of these electrons

will not penetrate the wall of the spacecraft, but could penetrate the

space suits worn during EVA, resulting in substantial doses to the

skin and eyes. Time-dependent factors are discussed in greater detail

in Appendix A.

Nuclear interactions of neutrons, protons and heavier nuclei with

spacecraft, space suits, Earth’s atmosphere, and the human body

produce secondary particles that add to the radiation field.

A more detailed summary of the near-Earth radiation environment

and the factors that modulate it can be found in NAS/NRC (2000)

and NCRP (2000a) and references therein.

Radiation monitoring instrumentation and measurement strate-

gies vary according to charge, particle type, energy, and where the

measurements are made. For purposes of this Report, we will classify

the radiation environment according to particle types and energies

and where the measurements are to be made, namely: outside the

spacecraft, inside the spacecraft, and inside EVA suits. The figures

in parentheses are the approximate particle ranges of interest in

energy (in MeV) and the unrestricted linear energy transfer (L�) in

water (in keV �m�1).

4.1 Trapped Electrons (0.5 to 6 MeV; �0.2 keV �m�1)

Trapped electrons with energies �0.5 MeV will penetrate some

parts of EVA suits. Electrons experience primarily electromagnetic

interactions. Because they are singly charged point particles and

not subject to strong nuclear interactions, secondary particle produc-

tion is limited to electrons, positrons and photons (through annihila-

tion and bremsstrahlung), which do not contribute substantially to

the radiation dose. Principal measurement locations are outside the

spacecraft and inside EVA suits. For aluminum shielding, primary

4.4 TRAPPED AND SOLAR PROTONS / 31

electrons are the dominant source of dose only for shields with areal

densities �0.15 g cm�2 (NCRP, 2000a). The relative importance of

electron and proton doses inside EVA suits is under study (Wilson

et al., 2001; Zeitlin, 20008).

4.2 Reentrant and Splash Albedo Electrons

(1 MeV to �1 GeV; 0.2 to �3 keV �m�1)

Reentrant electrons are the decay products of unstable nuclei pro-

duced by the nuclear interactions of trapped GCR ions. Splash albedo

electrons are scattered upward by interactions in the atmosphere.

The high energies of the parent nuclei boost these electrons to much

higher energies than trapped electrons. Badhwar et al. (2001)

recently made a detailed study of the contribution of reentrant and

albedo electrons at ISS altitude and inclination to the dose to the

bone marrow during EVA. They found the dose to be more than 10

times greater than the dose from trapped electrons. Moreover, the

hardness of the reentrant electron spectrum makes the dose from

those electrons rather insensitive to the addition of shielding, at

least at the level practical for an EVA suit. So while the trapped

electron dose will still dominate in thinly shielded locations, the

reentrant and splash electron dose dominates as the shielding thick-

ness is increased. The high energies of these electrons dictate that

measurements be made inside and outside the spacecraft, and inside

EVA suits.

4.3 Trapped Protons (�10 MeV; �5 keV �m�1)

Protons at energies �10 MeV (on the order of the nuclear binding

energy) will produce slow target fragments through compound

nucleus formation and decay. However, they will not penetrate space-

craft or EVA suits and therefore are not a measurement objective.

4.4 Trapped and Solar Protons and Light

Nuclear Particles (10 to 400 MeV; 0.3 to 5 keV �m�1)

Protons with energies greater than approximately 10 MeV will

penetrate EVA suits and spacecraft walls, and more energetic protons

8 Zeitlin, C.J. (2000). Personal communication (Lawrence Berkeley National Labora-

tory, Berkeley, California).


will penetrate into locations that are more heavily shielded, e.g., by

equipment racks or storage lockers. Nuclear interactions in materi-

als and tissue will produce low-energy, highly ionizing target frag-

ments and will knock out light nuclear particles (Z � 1, 2 and

neutrons) with velocities comparable to that of the projectile. There

is a small component of low-energy SPE light ions, but they are

thought to contribute a negligible amount to radiation risk (NCRP,

2000a). Radiation monitors and dosimeters sensitive to high-energy

SPE protons and light nuclear particles can be placed inside and

outside EVA suits and spacecraft.

4.5 Galactic Cosmic Radiation Ions and High-Energy

Secondary Fragments

(�50 MeV n�1; Z � 1; 1 to 1,000 keV �m�1)

The GCR ion spectrum peaks at energies around a few hundred

MeV n�1. Charged particles at these energies can penetrate into

the interior at many spacecraft locations and through substantial

thicknesses of tissue. Nuclear interactions will produce high- and

low-energy secondary charged particles, which are often more pene-

trating, including pions. The high-energy secondary particles arise

mainly from projectile fragmentation and will have velocities compa-

rable to that of the primary. Pions are created in nuclear interactions

above a few tens of million electron volts. The contribution of pions

to the dose is not known, but is under study at this writing (Wilson,

2000).9 Primary and secondary ions over a wide energy range will

be present at all locations of interest, both inside and outside EVA

suits and spacecraft.

4.6 Charged Target Fragments

(�10 MeV n�1; 2 to 1,200 keV �m�1)

Target fragments (i.e., pieces of the struck atomic nucleus) pro-

duced in nuclear interactions have very short range and therefore

deposit a large amount of energy very close to the location where

they are created. If this is within the human body, they can produce

substantial biological effects. The most probable energy is �5 MeV,

9 Wilson, J.W. (2000). Personal communication (Langley Research Center, National

Aeronautics and Space Administration, Hampton, Virginia).

4.8 SUMMARY FOR PARTICLE TYPES / 33

the average kinetic energy is 10 MeV, and the distribution falls to

one percent of the peak at 40 MeV (Shavers, 1999; Wilson et al.,

1988a; 1988b). Measurements of these particles would be needed

only inside space suits and spacecraft.

4.7 Neutrons (0.1 to 500 MeV)

Neutrons present in LEO are produced by nuclear interactions in

the upper atmosphere (albedo neutrons), and by interactions of high-

energy protons and heavier nuclei in shielding and tissue. The neu-

trons in turn interact with atomic nuclei to produce highly ionizing

charged secondary particles.

Measurements at high orbital inclination (57 degrees, slightly

higher than the 51.6 degrees for ISS) showed that at moderately

shielded locations most of the neutron H came from neutrons above

1 MeV (Badhwar et al., 1996a). Armstrong and Colborn (1992) calcu-

lated that for GCR-produced neutrons, almost half of the neutron H

is from neutrons above 10 MeV, and Armstrong recently calculated

that up to 20 percent of the total H on ISS from all particle types

will be from neutrons with energies �10 MeV (NASA, 2001).

A recent workshop considered the contribution of neutrons to radi-

ation dose to crews in LEO (NASA, 2001). The workshop concluded

that no single detector can cover the wide energy range of interest,

and that the detectors currently in use have not been adequately

calibrated. TLDs used on the Space Shuttle and ISS are ‘‘almost

completely insensitive to neutrons’’ and ‘‘measure absorbed dose and

not the biologically important dose equivalent.’’ Because of the high

penetrating power of neutrons, measurements are needed internal

and external to spacecraft, and inside EVA suits.

4.8 Summary for Particle Types

Table 4.1 summarizes the locations where radiation measure-

ments are needed as a function of the particle types discussed in

Section 4.1 through 4.7. The relative contributions to D and H from

each component (including the secondary radiation) at each location

will vary according to several factors, including the spacecraft inter-

nal mass distribution, the EVA suit design and materials, and the

site of interest within the human body. As a rule, D due to electrons

will be insignificant for all but the most thinly shielded locations of

an EVA suit.


TABLE 4.1—Summary of locations where radiation measurements

are needed, as a function of particle types, for operations in LEO.

EVA Suit Spacecraft Spacecraft

(internal) (external) (internal)

Trapped electrons X X —

Reentrant and albedo electrons X X X

Trapped protons (�10 MeV) — — —

Protons and light nuclear particles X X X

(�10 MeV)

GCR ions and secondary charged X X X

fragments

Charged target fragments X — X

Neutrons X X X

5. Approach to DoseAssessment forAstronauts

Dose assessment requires evaluation of specific quantities to man-

age exposures at the time of operation and for implementation of

career and short-term limits. The basic quantities to be evaluated

are given in Table 2.1 and the corresponding limitations are given

in Tables 2.3 and 2.4. The means by which dose is evaluated in

practice will be discussed in this Section, as well as the relationship

of radiation environment models, shielding models, and radiation

monitoring both inside the spacecraft and outside.

A basic quantity is mean absorbed dose in an organ or tissue (DT).

It is calculated from the distribution of D at points in the relevant

organ or tissue in terms of the L contributions at the specific points

[Di(L,x)], and averaged over the organ or tissue, as shown below.

The dose contribution from particles of type i with L at point x is:

Di(L,x) � � dt � �(x)�1 L &i[Ei(L),�,x,t] � dEi(L)/dL � d�, (5.1)

where &i[Ei(L),�,x,t] is the fluence of particle type i at the point x

at time t with energy Ei(L) (the multi-valued energy for given L of

a particle i) moving in direction �, L is the linear energy transfer

in appropriate units, � dEi(L)/dL � is the multi-branched derivative,

and �(x) is the tissue mass density.

DT is evaluated as the mean absorbed dose over all radiation com-

ponents i and organ tissue T as:

DT � MT�1 �

x

�(x) dx �i

� Di(L,x) dL, (5.2)

where MT is the total tissue mass.

The dose equivalent is defined at a point x in the tissue as:

H(x) � �i

� Qi[Li (E)] Di(L,x) dL, (5.3)

where Qi[Li(E)] is the quality factor for particle type i with Li(E)

(the multi-valued L for a given E of a particle i).

35

36 / 5. APPROACH TO DOSE ASSESSMENT FOR ASTRONAUTS

The organ dose equivalent (HT) is determined as an average of H

over all points in the organ or tissue as:

HT � MT�1 �

x

�(x) H(x) dx (5.4)

where the bar denotes the averaging process and T denotes the

organ or tissue over which the average is performed. Since, for space

radiations, HT is an acceptable approximation for HT (Section 2.2),

the effective dose (E) is then given as:

E � �T

wT HT � �T

wT HT. (5.5)

The evaluation of gray equivalent (GT) in a specific tissue of type T

is given in terms of the field-weighted quantities as:

GT � �i

Ri DT,i (5.6)

where DT,i is the mean absorbed dose to tissue T from the interior

field component i and Ri is the recommended value of the relative

biological effectiveness (i.e., a best estimate) for the specified particle

type i given in Table 2.2. Note that GT is not a point function and

leads to computational complexity. Note that the Ri values are energy

independent except those for neutrons (Table 2.2). A dose-averaged

value for neutrons (Rn,T) can be evaluated for specific tissues knowing

the neutron field and the appropriate tissue dose conversion factors

[Cn,T(E)] (ICRP, 1996; ICRU, 1998; Yoshizawa et al., 2000) as:

Rn,T �� dt � Rn(E) Cn,T(E) &n(E,�,x,t) dE d�

� dt � Cn,T(E) &n(E,�,x,t) dE d�, (5.7)

where &n(E,�,x,t) is the ambient neutron interior fluence at the

astronaut location in the spacecraft. The use of Rn,T reduces the

computational complexity but there is as yet insufficient RBE data

for most radiation types to evaluate these quantities for the full

interior radiation field. NCRP (2000a) has discussed the use of

conservative values for Ri components for which RBE values are

unknown (Table 2.2).

Figure 5.1 outlines the dose assessment scheme for use in LEO.

The knowledge required for evaluation of organ or tissue dose and

the dose-limit quantities for the astronaut includes the:

● radiation environment at the exterior surface of the spacecraft

● relationship of the exterior radiation environment to the interior

radiation environments of the Space Shuttle, ISS, and space suit

5. APPROACH TO DOSE ASSESSMENT FOR ASTRONAUTS / 37

Dose Records:

E, GT,H

T

Human Shielding Model

Linked personal records

External: I

Internal: D, H, D(L), f (y)

Personal: D, H, D(L)

Measurements Records:

Crew Personal Dosimetry

Measurements:

D, H, D(L)

D, ),)

Spacecraft External

Measurements:

, I

Occupancy

Factors

Internal Environment

STS/ISS/Spacesuit

Shielding models

Environmental Models:

Galactic Cosmic Rays

Albedo Neutrons

Trapped Particles

Solar Particle Events

List of quantities:

&i(E,�,t ): external or internal fluence; with particle type i, time t, energy E, direction �

I: ionization current in a tissue-equivalent ion chamber

D: absorbed dose, at a point

H: dose equivalent, at a point

D(L): absorbed dose distribution over L, at a point

f(y): frequency distribution of lineal energy

E: effective dose

GT: gray equivalent

HT: organ dose equivalent; approximation to equivalent dose (HT)

Fig. 5.1. Recommended dose assessment method.

● timeline of astronaut activity within those environments

● transmission of radiation from the internal radiation environ-

ment to specific organ tissues

● relationship of the organ exposure field to organ or tissue dose

● relationship of organ or tissue dose to the dose-limit quantities

Use of data measured during a mission relies on the knowledge

of a relation between the measured values and computed results.

This is accomplished with an adequate understanding of the

response function of the measurement device to predict the response

from the expected particle field. Adjustments in the estimated fields

from calculations are obtained by comparing predictions and actual

measured results, and this will then allow improvements in radiation


environment characterizations and dose estimates. In this respect,

the correction to the interior fields associated with the less known

trapped radiations will be a prime target of such procedures.

5.1 Exterior Exposure Field

5.1.1 Radiation Environment Models

In practice, the external radiation environment in LEO is repre-

sented by various predictive models and descriptive fits. Best known

are GCRs (Badhwar and O’Neill, 1992; Davis et al., 2001) and their

associated quiet-time geomagnetic transmission factors. Periods of

geomagnetic storms provide only small corrections to the long-term

averaged fields in LEO. The albedo neutrons produced by the impact

of GCR with Earth’s atmosphere are also well known (Korff et al.,

1979; Lockwood, 1972; Preszler et al., 1972). GCR models are well

developed and the predictive model of Badhwar and O’Neill (1992)

gives a reasonable representation of the local radiation environment

outside Earth’s geomagnetic field. More recent descriptive fits associ-

ated with the Advanced Composition Explorer spacecraft project are

now available and have the advantage of validation by measure-

ments with high-quality instrumentation in the L1 libration point

(i.e., about 1.5 million kilometers from Earth in the sun’s direction)

where Earth and solar gravitational strengths balance each other

(Davis et al., 2001).

The current trapped radiation models are uncertain by a factor of

two for geomagnetic quiet times and the trapped radiation experi-

ences large fluctuations during geomagnetic storms for which there

are no predictive models. In addition to the uncertainty in intensity,

the trapped radiations are highly directional and their velocity vec-

tors are constrained to lie within several degrees of a plane normal

to the local geomagnetic field line. The AP-8 (Aerospace Proton Envi-

ronment, Version 8) model for protons and AE-8 (Aerospace Electron

Environment, Version 8) model for electrons (Vette, 1991) represent-

ing the trapped radiation environment are the basic standard, based

on data obtained by NASA and Air Force measurements from 1958

to 1970. The AP-8 model includes angular dependence and both

AP-8 and AE-8 provide the omnidirectional fluence estimates for

geomagnetic quiet conditions in current usage.

SPEs occur mainly during times of elevated solar activity and

are unpredictable in magnitude and exact time of occurrence. Their

detection must rely on satellite measurements available in real-time

5.2 INTERIOR EXPOSURE FIELD / 39

from some space platforms. Transmission through the geomagnetic

field to LEO is normally represented by the vertical cutoff model

derived transmission factors for 20 km height (Shea et al., 1988) and

scaled in altitude according to a magnetic dipole approximation.

Recent tables of global vertical cutoff transmission factors have not

been published.

5.1.2 Spacecraft External Measurements

External measurements can be made of the outside radiation envi-

ronment to allow correction to the models to reduce the uncertainties.

An ion chamber with well-defined wall thickness can serve to monitor

the short-term variations in the electron environment for EVA opera-

tions and particle spectrometers provide useful data to reduce the

uncertainty in the trapped particle fields and during SPEs. Recom-

mended instrumentation is discussed in Section 6. The principal

external measurements are: &i(E,�,t), the external fluence for parti-

cles of type i at time t with energy E moving in direction �; and I,

the ionization current in a tissue-equivalent ion chamber.

5.2 Interior Exposure Field

5.2.1 Shielding Models for the Space Shuttle, International

Space Station, and Space Suits

Shielding models are required for the Space Shuttle, ISS, and

space suits to allow evaluation of the interior radiation environment

to which the astronaut is exposed. The models consist of the distribu-

tion of materials about each field point within the vehicle and appro-

priate computational procedures to evaluate the interior field of

transmitted particles. The internal radiation environment generally

shows high directionality with large spatial gradients and varies

over time with both short- and long-term temporal scales. The distri-

bution of materials within the three structures (Space Shuttle, ISS

and space suit) is represented with computer generated solid geome-

try models. The Space Shuttle model (Atwell et al., 1989) is assumed

to be made totally of aluminum alloy 2219. Higher fidelity ISS and

space suit models allowing a distribution of materials are desirable.

With the exception of the absolute intensity of the trapped radia-

tion environment and possible SPEs, the interior radiation environ-

ment can be evaluated with high-speed computational models as


has been demonstrated for Space Shuttle flight measurements

(Badhwar et al., 1995a; Shinn et al., 1998). High-speed computational

procedures allow rapid mapping of the interiors of the Space Shuttle

and ISS. The computational procedures are outlined in Appendix B.

Fast computational procedures for charged ions and neutrons are

well developed (Clowdsley et al., 2000; Cucinotta et al., 1997; Wilson

et al., 1995). Fast computational procedures of electron and photon

transport are available for limited shielding materials (Seltzer, 1980)

and more general codes are still being developed. This series of

electron and photon transport codes requires the approximation of

shielding (including tissue) to be taken as equivalent amounts of

aluminum and only provides dose, with no information on electron

and photon spectra.

5.2.2 Spacecraft Internal Measurements

Selectively located active and passive instrumentation allows fur-

ther adjustments to the evaluated internal radiation environment,

thus reducing uncertainty. Particle spectrometers allow some evalu-

ation of transmission factors and dosimetric devices allow evaluation

of calculated dosimetric quantities. Both active and passive dosimet-

ric devices are useful for evaluation of D, D(L), the frequency distri-

bution of lineal energy [ f (y)], and H, as discussed in Appendix A.

Personal dosimeters worn by astronauts can determine some or all

of these quantities in adjacent tissues.

The interior radiation environments of the Space Shuttle and ISS

will be monitored by various instruments, and the measurements

can be used to adjust the modeled trapped particle intensity, thereby

reducing the uncertainty in estimates of the interior radiation envi-

ronment. There are long-term variations depending on solar activity,

which modify the environmental spectra as well as intensity, so

that spectrum-sensitive instrumentation is desirable. Furthermore,

active instrumentation is desirable to allow detection of time-

resolved trapped and GCR components. Whatever instrumentation

is used within the Space Shuttle or ISS, the instrument response

needs to be accurately defined for direct comparison of measurement

with predicted response as the basis for modifying the calculated

results to reduce uncertainty in estimates of the interior radiation

environment. The principal internal measurements are: &i(E,�,t)

5.3 TISSUE EXPOSURE FIELDS / 41

(the internal fluence for particles of type i at time t with energy E

moving in direction �), D, D(L), f(y), and H.

5.3 Tissue Exposure Fields

5.3.1 Human Shielding Models

Shielding models for astronaut tissues are required to evaluate

the fields present at specific sensitive tissues. The models consist of

the distribution of the specific sensitive organs in the body and the

surrounding distribution of the astronaut tissues about each point

in that organ and computational models to evaluate the transmitted

fields. The presence of high directionality and large spatial gradients

coupled to the mobility of the astronaut poses a challenge to evalua-

tion of the exposure fields at specific tissue points. In addition, astro-

naut shielding models allow the evaluation of the effect of the

presence of the astronauts in the radiation environment on the local

interior field and in particular on the exposure of the personal dosim-

eter, and on the relationships between D and H at points on the

surface of the body and these quantities in deeper-lying organs. The

evaluation of organ or tissue dose for the astronauts is performed

with computerized anatomical models for both a male and a female

(Atwell et al., 1993). The models provide the distribution of tissues

with relatively high fidelity. A male data set derived from computed

tomography scans has also been developed but is not fully utilized

in protection practice (Qualls and Boykin, 1997).

5.3.2 Occupancy Factors

Occupancy factors keyed to known individual activity can be used

to estimate an astronaut’s exposures and can be corrected by the

personal dosimeter measurements. This is especially true during

EVA where large fluctuations in the trapped electron environment

or SPEs could greatly alter the dose gradients within the astronaut

tissues. In addition, one could have a combination of shielded and

unshielded TLDs in a personal dosimeter such that the shielded

dosimeters did not record trapped electrons during EVA.

6. Data Collection andInterpretation forDose Assessment

6.1 Introduction

Data collection and interpretation are essential components of

an operational radiation safety program for space activities. Mea-

surements from onboard detectors in combination with radiation

transport calculations should be designed to provide sufficient

information to:

● assess career exposures

● assess short-term exposures

● manage the exposures during all mission segments

There are several purposes of in-flight measurements. These

include the provision of field data (fluence, dose) both integral and

differential (with respect to time, LET, energy or direction, as appro-

priate), normalization of the calculations of the radiation field compo-

nents, and, in the case of personal dosimeters, the determination of

D and H. The quantities D and H refer to determinations averaged

over the sensitive volume of the detector as measures of D and H to

a point in the adjacent tissue.

In order to perform these functions, instrumentation must be able

to accommodate the complex nature of the radiation environment

in space and a range of mission profiles. Additional complications

are introduced by the variety of astronaut activities both inside and

outside the spacecraft.

Several types of instrumentation are required. Area monitors,

fixed or portable, will be used to measure fluence and fluence rate,

D and absorbed dose rate (D·), from which H and dose equivalent

rate (H·) may be calculated. Due to the complex nature of the radiation

field it may be necessary to differentiate fluence and D with respect to

particle type and energy, LET, and direction. Active instrumentation

should have a time resolution sufficient to identify temporal varia-

tions in the radiation field. These temporal variations may be associ-

ated with the characteristics of the orbit or other external factors.

42

6.2 DOSE QUANTITIES TO BE DETERMINED / 43

Personal dosimeters can be used to make measurements of D or H

to tissue at a point on the surface of the body and assign these

directly to individuals. In addition, alarm or warning instruments

can be used to support in-flight implementation of dose management

and ALARA actions.

6.2 Dose Quantities to be Determined

The objective of the adopted dose assessment approach (Section 5)

is to determine the necessary radiation protection quantities for the

limitation of stochastic and deterministic effects. For stochastic

effects this quantity is E determined from E � �T

wT HT, where HT

is used as the surrogate for HT (Section 2.2.2). For deterministic

effects the quantity is GT determined from GT � Ri DT, where DT is

weighted by Ri for the particle type (Section 2.2.1). Passive personal

dosimeters can be used to determine D and H to a point in adjacent

tissues. The quantities DT and HT to organs and tissues, and thus

E and GT are determined from a combination of radiation transport

computations and dosimetry measurements (Section 5).

Thus, an appropriate measurement package should provide:

● input data for the radiation transport calculations. The combina-

tion of calculations and measurements will yield estimates of E

and GT ;

● measurement of desired point quantities such as D, the spectrum

D(L), and H to adjacent tissues; and

● measurements in support of in-flight dose management and

ALARA actions.

Calculations serve a dual purpose. Firstly, they are essential to

mission planning in that they provide preflight estimates of the doses

that would be received by the astronauts during the different phases

of the proposed mission. Secondly, they are an essential component

of the dose assessment process for determining the actual doses

received by the astronauts during the mission. In the latter role,

the calculations are combined with measurements from area and

personal dosimeters in order to arrive at the final values of HT, E

and GT.

As described in Section 5 (Figure 5.1), data from the area monitors,

both internal and external to the spacecraft, are used to determine

essential field quantities as input to the calculations in order to define

the internal radiation environment at different locations inside the

space vehicle. Data from the measurement package [i.e., D, D(L),

44 / 6. DATA COLLECTION AND INTERPRETATION

and H] are then used along with the calculations to provide estimates

of HT, E, and GT. As already described, this can be achieved by using

the measured D, D(L), and H to normalize the computations before

proceeding to compute values for each astronaut. Alternatively, the

computations may provide correction factors for each astronaut to

convert H values determined from personal dosimeters into HT, E

and GT values. In either approach, it is a combination of radiation

transport computations and dosimetry measurements that is used

to provide the ultimate values for entering into the record.

The measurement packages described in this Section have suffi-

cient flexibility to support either approach. The proposed packages

consist of:

● active devices, including TEPCs and particle spectrometers in

strategic locations within the spacecraft, to collect data for short-

term and career records;

● passive devices, including TLDs and PNTDs as personal dosime-

ters for the crew; and

● active devices to act as alerts for high transient doses and dose

rates.

In this Section potential alternative, or additional, dosimetry

devices that may be employed in the future are discussed, including

solid-state active personal dosimeters, solid-state ionization cham-

bers, and high-intensity electron field monitors.

6.3 Proposed Measurement Package

6.3.1 General Discussion

The measurement package proposed by NCRP follows naturally from

the determination of the observables to be measured (Section 6.2)

and the capabilities of available and planned instrumentation. No

single device can determine the required dose quantities for all com-

ponents of the radiation field. Since there are large variations in the

magnitudes of the contributions to total D and H from the different

particle types it is not generally possible to determine D or H from

just one component (or components) and apply a correction factor to

determine D or H for the whole field. Therefore, practical measure-

ment and dose assessment considerations suggest the need to evalu-

ate three main components of energy deposition, namely: (1) low-

LET charged particles and photons; (2) neutrons, and protons that

undergo nuclear interactions which produce high-LET secondary

6.3 PROPOSED MEASUREMENT PACKAGE / 45

particles; and (3) energetic charged nuclear fragments with Z � 1

(HZE particles).

The rationale for the recommended measurement package is to

ensure that suitable measurements are made in each of the above

three categories, that will allow determination of the dose-limit

quantities, without unnecessary overlap or duplication that might

result in ‘‘double counting.’’ The choice of measurement device or

devices is dictated by radiation response characteristics (particle

type, particle energy, quantity to be determined), operational charac-

teristics (direct determination of dose quantities, input to model

calculations, desired accuracy, dose management, and ALARA sup-

port), as well as practical issues such as reliability, robustness and

availability.

An additional important consideration is the time required for

analysis of the devices compared with the duration of the mission.

Long duration ISS missions may place constraints upon the fre-

quency of analysis of passive dosimeters. If the personal measure-

ment package on ISS does not include active dosimeters, facilities

to read passive devices onboard the spacecraft may be necessary to

fully implement dose management and ALARA. Such facilities may

not be necessary on the Space Shuttle. Furthermore, the proposed

passive personal dosimeter package is to be used both when the

astronaut is inside the spacecraft and during EVA. However, supple-

mental passive personal dosimeters may be needed for additional

locations on the astronaut during EVA because of spatial variations

in the degree of shielding provided by the space suit.

The addition of TEPC (Section 6.3.2.1) to routine area dosimetric

evaluation in the mid-1990s provided direct information on energy

deposition in tissue by both high- and low-LET particles. The data

provided are in terms of D in a small mass of tissue collected in real-

time, and can be used to determine D·. Energy deposition events can

be combined to give distributions in lineal energy (y) that are related

to LET of the incident radiation. It may also be possible to use this

information to estimate unrestricted LET (L�) distributions, which

are required for determination of Q and thus H to a small mass

of tissue. Radiation transport models have been used to predict y

distributions measured by TEPCs (Badhwar et al., 1996c; Shinn

et al., 1999). From these calculations estimates of the radiation envi-

ronment within the space vehicle have been made. The success of

this approach relies upon an accurate estimate of the external radia-

tion field, the geometry and composition of the spacecraft, and the

response of TEPC detectors to HZE particles. In general, TEPCs

have little dependence of response on angle of incidence, that is,

their response is approximately isotropic.


Particle spectrometers (Section 6.3.2.2) can identify particles individ-ually by charge and energy. This is necessary information when consid-ering radiation effects that are sensitive to particle type and fluence,as opposed to D and LET. It is also possible to make comparisonsbetween the data and calculations. Particle spectrometers generallyhave limited angular acceptance and energy, and poor statistics forZ � 2. Details of properties and performance of both TEPCs and parti-cle spectrometers can be found in Appendices B and C.

Thermoluminescent dosimeters (of type TLD-100� or TLD-700�)have been used routinely for both passive personal dosimeters andpassive area dosimeters since the Skylab mission. In addition toTLD, high-LET passive dosimeters have been flown on almost allmissions. TLDs record the energy deposited for photons, electrons,positrons, pions, protons and heavier charged particles. The responseis dependent upon particle type and energy. For example, whencalibrated in terms of tissue kerma to high-energy photons, TLD-100� and TLD-700� overestimate D in tissue (by several percent)for high-energy protons and begin to underestimate D in tissue forparticles with LET (in water) above approximately 10 keV �m�1.The underestimation has been shown to be approximately 50 percentat 100 keV �m�1. However, the exact values of the over- and under-response for a given LET and particle type are not well established.The values vary from material to material and, for a given TLDmaterial, values vary in the published literature. Furthermore, TLDsprovide a single number related to D and provide limited spectro-scopic information on the quality of the radiation field, i.e., D(L).Because of these properties, TLDs alone do not provide sufficientinformation to determine H. Furthermore, the design of the TLDpackage may introduce an unwanted dependence of response onangle of incidence. Details of TLD properties and performance canbe found in Appendix C.

PNTDs have been a standard component in crew and area dosime-try packages. These detectors respond to high-LET particles, includ-ing secondary charged particles from neutron interactions, but areinsensitive to photons, electrons and other charged particles withLET (in water) values less than approximately 10 keV �m�1, depend-ing upon PNTD material. Processing of PNTD can yield LET distri-butions for the radiation field, as well as particle type and energy.PNTDs have a strong dependence of response on angle of incidence.Details of the properties and performance of PNTDs can be foundin Appendix D.

6.3.2 Proposed Measurement Package: Active Devices

Active devices record and display data in real-time, or near real-time. For most applications both the readout time and time resolution


are significantly shorter than the time for Space Shuttle or ISS to

complete one orbit. This makes it possible to obtain measurements of

rates as well as time integrated values. In most cases the time resolu-

tion is sufficient to select data from defined portions of the orbit (e.g.,

when traversing SAA). Most active instruments can be turned on or

off so as to operate in specific circumstances and conditions (EVA,

SPE). Active detectors require electrical power that can be provided

through connections with power supplies in the spacecraft or through

batteries. They can be used as fixed or portable area monitors and

devices are under development that are sufficiently small to be carried

on an astronaut’s person. The types and characteristics of detectors

to be used in LEO should be determined by the requirements of dose

assessment, as specified in Section 5. The following detectors can

provide sufficient data to fulfill those recommendations.

6.3.2.1 Tissue Equivalent Proportional Counters. A TEPC is an

active detector that is designed to measure energy deposition in

volumes of tissue comparable to the dimensions of the nuclei of

mammalian cells. Data are recorded on an event-by-event basis such

that one can obtain a distribution of biologically relevant energy

deposition events. The wall and gas cavity can be designed to make

a TEPC sensitive to incident photons, neutrons and charged particles

from protons to heavy ions.

When the data are integrated over the complete distribution of y,

TEPCs can generate D and D·. The frequency distribution of lineal

energy [ f (y)] depends on characteristics of the radiation field

[&i(E,�,t)] (i.e., the distribution of fluence for particle type i as a

function of energy E, direction �, and time t) and the response of

the detector. Therefore, f(y) can serve as a test for radiation transport

models. It can also be used to obtain D in terms of lineal energy

[D(y)]. This is not a replicate of D(L), but can be used to obtain an

approximation to Q for protons and heavier particles. The detector

has a large dynamic range such that it can identify signals from

charged particles that deposit energy from 0.2 to 800 keV in the

sensitive volume. It is also sensitive to neutrons and this contri-

bution is also included in D(y). Data from a TEPC can be displayed

continuously and also stored for later transmission to the Mission

Control Center.

6.3.2.2 Solid-State Detectors. Solid-state detectors record the

energy deposited by a charged particle. The ratio of the deposited

energy (dE) to the thickness of the detector (dx) yields the approxi-

mate L� for the incident particle. Thus, a single detector can provide

an estimate of &(L,t) [i.e., the distribution of fluence as a function


of LET (L) and time (t)] that can be integrated to yield D and D·

for

protons and heavier charged particles. It can be fabricated into a

compact detector for use as a portable area monitor or personal dose

rate meter with on-demand readout (Badhwar, 2000).

Often several of these detectors are combined to form a particle

telescope that measures both energy deposition and position. When

these data are combined, this detector yields a more accurate esti-

mate of L� and thus &(L,�,t) (i.e., the distribution of fluence as a

function of L, � and t). It can also be used to obtain D(L) and Q for

heavy particles, but with the restriction that the incident particles

originate from a fixed direction depending on the orientation of the

detector. Because of size limitations this type of detector is sensitive

over a restricted solid angle (d�).

The energy lost in one or more solid-state detectors can be used

to identify the particle charge (Z) as well as the incident energy (E)

to ultimately obtain &i(E,Z,t) (i.e., the distribution of fluence as a

function of E, Z and t). Several of these detectors can be combined

to point in different directions to provide a more complete description

of the radiation field either outside or inside of the spacecraft.

6.3.2.3 Active Electronic Personal Dosimeters. There are a num-

ber of real-time electronic personal dosimeters available, mostly

using silicon-based detectors (Ortega et al., 2001; Texier et al., 2001).

Most of these have been designed to measure photon and beta radia-

tion, and might be considered for the measurement of the low-LET

component of the fields in spacecraft. Several have tissue-equivalent

encapsulation. However, even if used only to determine the low-

LET component, a full characterization of the charged-particle and

neutron response is necessary before any can be recommended. Neu-

tron and total field electronic personal dosimeters are not as fully

developed as are those designed to determine photon and beta radia-

tion fields. Direct ion storage dosimeters, described in Section 6.3.3.2,

can be operated as real-time devices to determine the low-LET

component.

6.3.2.4 Active Detectors for Electrons. Active detectors should be

specially configured to measure low-LET radiations, in particular,

electrons. This is normally not an issue inside the spacecraft but

could be of concern during EVA, since electrons above a few hundred

kiloelectron volts can penetrate the space suits. Since the trapped

electron intensity can change by many orders of magnitude during

and following a large magnetic storm, due to short-term perturba-

tions of the geomagnetic field, it is recommended that an active

detector sensitive to electrons be installed outside of the spacecraft


to serve as a monitor for fluctuations in the electron component of

the space radiation environment for dose management during EVAs.

Such a monitor could be a simple ionization chamber with a wall

thickness sufficient to attenuate very low-energy electrons but thin

enough to record electrons that could penetrate a space suit. The

output would be the ionization current (I). Because the trapped elec-

trons occur intermittently, it will be necessary to monitor or suppress

the signal from background radiation that occurs continuously.

6.3.3 Proposed Measurement Package: Passive Devices

Passive devices can be used as both area and personal dosimeters.

Since no single passive device is capable of dose measurement across

the full spectrum of LET, measurement packages should be designed

to have optimum performance in the three different regions noted

above, namely: (1) low-LET charged particles and photons,

(2) neutrons and protons that undergo nuclear interactions, and

(3) HZE particles. Passive dosimeters currently in use on LEO Space

Shuttle flights are TLDs, specifically TLD-100� or TLD-700�, to

obtain D in tissue (in gray) calibrated against 137Cs gamma rays.

That is, the 137Cs gamma-ray dose in tissue (or kerma in tissue)

required to give the same thermoluminescent (TL) response in TLD

as the space irradiation is reported as D in tissue at the location of

the TLD detector. However, TLD-100� and TLD-700� are not the

best TLD materials to use, since they have a variable and decreasing

response at high-LET. Other TLD materials may be better, in that

they may be more suited for use at low-LET (i.e., �10 keV �m�1)

with very little sensitivity to higher-LET particles. Furthermore,

since there may be a need for some passive elements to be read out

frequently onboard, and others that are read only on return to Earth,

the mode of readout of the luminescence dosimeters should be care-

fully considered. OSLDs and electronic dosimeters may be viable

options. However, any new TLD or OSLD material, or new electronic

dosimeters, will need to be fully characterized for the space radiation

environment.

To read the high-LET component of the radiation field PNTDs can

be employed. These are generally insensitive below a LET in water

of approximately 10 keV �m�1 (depending on material) and may be

employed in a mode that provides D, D(L) (or D averaged over certain

ranges of L), and H.

The various elements of the proposed packages are discussed

below.


6.3.3.1 Low Linear Energy Transfer Dosimetry: Thermolumines-

cent Dosimeters. As noted, after calculation of the equivalent 137Cs

dose the TLD response needs to be corrected for the variation in the

efficiency of TLD as a function of particle LET and for the neutron

response of the dosimeter. This would require taking into account

the best estimates of the radiation field components at an appropri-

ate spacecraft location. Therefore, it is recommended that passive

dosimetry using TLDs or OSLDs be limited to the determination of

D from the low-LET charged particle and photon component only

(i.e., for L � 10 keV �m�1). This can be achieved by choosing a

device (a luminescence dosimeter, i.e., a TLD or an OSLD) that has

a minimal response to the other high-LET components. Dosimeters

LiF:Mg,Cu,P (for TLD) and Al2O3:C (for OSLD) should be considered

since they are predicted to be very sensitive to low-LET radiation,

and very insensitive to high-LET radiation. This prediction is based

on their gamma-ray dose response, which shows a considerably lower

saturation dose for the former materials compared with, say, TLD-

100�. For 5 MeV alpha particles this prediction has been shown to

be true (Appendix C). Lithium fluoride (LiF) is a tissue-equivalent

material for low-LET radiations and for photons. Al2O3:C is less

tissue equivalent but still acceptable. One advantage of switching

to OSLDs would be the possibility of providing small, lightweight,

low-power OSL readers for the spacecraft so that the astronauts

could read their own dosimeters during long-duration flights. In

either case, these materials may provide excellent discrimination

between low- and high-LET radiation.

As an alternative to OSLDs, one could consider radiophotolumi-

nescence (RPL) glasses (Piesch et al., 1990; 1993). Like OSLDs, these

are optical-readout dosimeters, but they provide a permanent inte-

gration of the dose and cannot be re-used. Readout equipment is

similar to that of OSLDs. However, like OSLDs, RPL glass dosime-

ters have not been tested for their response to particle radiation

over a wide range of LET. Thus, before adoption of OSLDs, or RPL

dosimeters, research is required to fully characterize their response

in detail.

CaF2:Tm dosimeters (such as TLD-300�) can provide useful quali-

tative information about the particle LET spectrum using the peak-

height-ratio method (Appendix C). It has been suggested that the

variation of the peak height ratio with LET for either LiF:Mg,Ti-

based detectors (e.g., TLD-100� or TLD-700�) or CaF2:Tm detectors

(e.g., TLD-300�) could be used to provide information about the

quality of the incident radiation field by calculating a ‘‘TLD average

Q,’’ namely a parameter QTLD (Appendix C). However, the variation

of the peak-height-ratio with LET is nonlinear. Furthermore, it is


unclear what the obtained value of QTLD means. In general, where

the radiation field has a significant neutron component, QTLD will

differ from Q for tissue because of the lower neutron kerma and

absorbed dose in CaF2:Tm than in tissue. Despite claims by some

authors that the QTLD value matches well the ‘‘true’’ average value

for the low- and high-LET components, comparisons of QTLD with QD

(D averaged Q) calculated from the TEPC data differ by as much as

a factor of two. Thus, we recommend the use of CaF2:Tm only as

additional input into the model calculations and not as a quantitative

method to obtain H for low- plus high-LET components.

6.3.3.2 Direct Ion Storage Dosimeters. In addition to personal

dose rate meters discussed in Section 6.2.2.3 there are new solid-

state detectors that combine an ionization chamber with a semicon-

ductor. Specifically, the direct ion storage dosimeter is based on

coupling a gas-filled ion chamber with a semiconductor nonvolatile

memory cell (Wernli and Kahilainen, 2001). These are compact integ-

rating devices that can be read out periodically and used to estimate

accumulated doses over periods of several hours to at least a year.

6.3.3.3 Neutron and High Atomic Number, High-Energy Particle

Dosimetry: Plastic Nuclear Track Detectors. To determine the neu-

tron and HZE components, a separate device (comprising one etched

track detector, or two or more such detectors of different response

characteristics) should be used (see Appendix D for descriptions of

methods and for references). From such a dosimeter, the point values

for H in adjacent tissues can be determined. Materials that can be

used are PADC/CR-39�, cellulose nitrate, and polycarbonate [Lexan

(General Electric Company Corporation, New York); Makrofol�

(Bayer Aktiengesellschaft Corporation, Leverkasen-Bayerwerk,

Germany)]. Herein we refer to all such materials as PNTDs.

The LET (in water) threshold of PADC/CR-39� is in the range 5

to 10 keV �m�1, enabling protons of energy up to about 10 MeV to

be detected directly, as well as HZE particles. Secondary charged

particles from nuclear interactions of higher energy protons and of

neutrons, in the detector material itself or surrounding material,

are also detected. Full etch track analysis of stacks of PNTDs allows

the determination of HZE particle fluence and fluence rate, but is

very time consuming. By a measurement of the distribution of D

(fluence times LET) as a function of LET, H can be determined for

energy deposition by all particles in the (approximate) LET range

of 5 to 1,000 keV �m�1. This can be done using single detectors, but

is still time consuming. Automated analysis systems have been used

but further development is needed. Another, more approximate


approach, is to use a series of different materials each of different

LET threshold to obtain particle fluence in a number of LET ‘‘bins.’’ In

both approaches, although LET calibration of the detector response is

in terms of LET in water, the numbers and types of secondary parti-

cles generated, for example by the neutron and higher energy proton

components, will be different than those generated in tissue. There-

fore, the determination of H (to tissue) will be approximate. PNTDs

may also be used to estimate the higher-LET component of GCR

using the simpler, less time consuming techniques developed for

neutron personal dosimetry. Electrochemically etched pits are iden-

tified and counted. Readout procedures are fully automated. HZE

particles will also produce an etchable track and be counted as if

produced by a neutron, or neutron-like iterations of high-energy

protons, and the same response factor applied (about 10 �Sv per

track for a typical electrochemical etch system), which will over-

estimate the HZE component of H. However, an additional chemical

etch allows discrimination. The response of PNTDs is angle depen-

dent and, in general, the interpretation of results requires assump-

tions with respect to the direction distribution of the field to which

PNTDs were exposed.

6.3.3.4 Use of Thermoluminescent Dosimeters and Plastic Nuclear

Track Detectors to Estimate Effective Dose. Measurements of H

with TLDs and PNTDs located at the surface and HT with TLDs and

PNTDs located in the organs of an anthropomorphic phantom were

made during a Space Shuttle mission to Mir (Yasuda et al., 2000).

The results indicate that the determination of H at the body surface

may be able to provide an assessment of the effective dose (E) for

the radiation field inside ISS. However, this conclusion needs to be

supported by further investigations for other shielding configura-

tions. The potential for developing a set of conversion coefficients

that directly relate H at the surface to E for the space radiation

environment, similar in concept to those used in other occupational

radiation environments (ICRP, 1996; ICRU, 1998), would be worth

investigating.

6.3.3.5 Superheated Drop/Bubble Dosimeters. Superheated drop/

bubble neutron personal dosimeters are available, and have been

flown on Mir and the Space Shuttle (Ing, 2001). These dosimeters

respond to thermal neutrons and fast neutrons upwards from about

100 keV, but require a normalization of the neutron response to the

neutron spectrum in spacecraft and corrections for proton and other

ion-induced bubbles. The use of this type of device must take into

account dynamic range limitations, the temperature dependence of


response or the efficiency of the correction applied for this effect,

and the possibility of false readings due to microphonics.

6.3.4 Recommendations for Measurement Packages

6.3.4.1 Recommendations for Area Monitoring. Active detectors

play an important role in terms of characterizing the radiation fields

inside and outside of the spacecraft. Detectors with time resolutions

on the order of minutes can be used to distinguish regular changes

in radiation during portions of the orbit (i.e., trapped radiation in

polar regions and SAA, galactic cosmic rays) as well as identify other

intermittent changes due to solar activity and magnetic storms.

A TEPC can provide dose and dose rate measurements with suffi-

cient time resolution. It is sensitive to photons, protons, helium, HZE

particles, and neutrons. It has the additional capability of providing

information on f (y) that can be used to obtain D(y). Although D(y)

is not identical to D(L), dose-averaged values of y have values very

similar to those for L and thus can be used to estimate Q (Appen-

dix A). A TEPC can be placed at fixed locations within the spacecraft

or can be operated as a survey meter if necessary.

Active solid-state detectors can provide a direct measurement of

particle fluence for protons and heavier particles in terms of &(L)

(i.e., the distribution of fluence as a function of L) for a single detector

or &(L,�) (i.e., the distribution of fluence as a function of L and �)

and &(Z,E,�) (i.e., the distribution of fluence as a function of Z, E

and �) using multiple detectors. These data can serve as calibration

points for particle transport codes used for dose assessment. When

integrated, they also provide an estimate of D to silicon that can be

converted to D to tissue. Because they estimate LET, they can also

provide estimates of Q. Active solid-state detectors can be used as

area monitors inside and outside of the spacecraft. A compact porta-

ble version has been developed and could be used as a personal

dosimeter with on-demand readout.

An active detector sensitive to electrons should be installed outside

of the spacecraft to serve as a monitor for fluctuations in the electron

component of the space radiation environment that can change by

many orders of magnitude during and following an SPE due to short-

term perturbations of the geomagnetic field. This could be of serious

concern during an EVA since electrons above a few hundred kiloelec-

tron volts can penetrate the space suits. Such a monitor could be a

simple ionization chamber with a wall thickness sufficient to attenu-

ate very low-energy electrons but thin enough to record electrons

that could penetrate a space suit.


The deployment of active instruments can be usefully supple-

mented by passive devices, for example, by PNTD stacks to deter-

mine both HZE particle fluence rates and LET distributions, and by

TLDs or OSLDs to examine the spatial distribution of the low-LET

component.

6.3.4.2 Recommendations for Personal Dosimetry. A measure-

ment package consisting of a TLD material for measurement of the

low-LET component and a PNTD stack to determine the high-LET

component would appear to provide the best current solution for

passive personal dosimetry in the complex radiation field experi-

enced in space. For TLD, LiF:Mg,Cu,P would appear to be an attrac-

tive material. Alternatively, Al2O3:C is the best currently available

OSLD material. To measure H from the high-LET components, the

use of PADC/CR-39� etched track detectors is proposed. Such devices

have been used as a part of the area monitoring or personal dosimeter

packages on the Space Shuttle, but are not currently planned for

ISS. It is recommended that they be used as personal dosimeters on

both vehicles. CaF2:Tm (e.g., TLD-300�) could also be used as an

adjunct personal dosimeter to provide additional information for

model normalization purposes, but not for quantitative determina-

tion of dose quantities.

With this measurement package for passive personal dosimetry,

H at a point in adjacent tissue is then obtained from:

H � DTLD/OSLD � � DPNTD(L) Q(L) dL (6.1)

where DTLD/OSLD is absorded dose recorded by TLD or OSLD in the

low-LET region (L � 10 keV �m�1) for which Q � 1. DPNTD(L) is the

D distribution in LET determined by PNTD over the high-LET range

(L � 10 keV �m�1) and for which Q is dependent upon L. Correction

may be needed for any overlap of the two responses so that intermedi-

ate-LET components are not double-counted.

It is recognized that this recommendation requires verification of

LET dependence of the TLD response of LiF:Mg,Cu,P and of the

OSLD response of Al2O3:C. Until such time as these data are avail-

able, therefore, LiF:Mg,Ti dosimeters (e.g., TLD-100� or TLD-700�)

may still be used. If so, however, this should still be along with

PNTDs, as part of the passive personal dosimetry package, in order

to provide LET data suitable for correcting the TLD dose response

for the L � 10 keV �m�1 component, and for estimating H for this

component from the PNTD results.

If PNTDs cannot be flown, a different approach has to be adopted.

Data needs to be obtained from a TEPC or particle spectrometer

used as a passive area dosimeter. If appropriate LET information

6.4 ACCURACY, PERFORMANCE TESTING, AND CALIBRATION / 55

is available from the area dosimeter, then correction of the TLD

response may be made by extracting (i.e., from TEPC data) a dose-

averaged LET (LD), from which one can evaluate a mean Q (Q) from

the known Q(L) relationship, and an effective efficiency (�) from a

measured �(L) relationship (Appendix C).

The value of H to the adjacent tissue is then approximated by:

H �Q D

�, (6.2)

where D is the dose of 137Cs or 60Co that is found to give the same

TL response from TLD as the space irradiation. This approach suffers

from the drawback that DTLD for the neutron component of the fields

will, in general, be significantly less than D in tissue for this compo-

nent. The Q derived from TEPC measurements should be applied

to D from all components to obtain the total H. The degree of approxi-

mation resulting from the adoption of the approach described will

depend on the proportions of the different field components and

therefore location in the spacecraft. Some new designs of passive

electronic dosimeters function as small tissue equivalent ionization

chambers. For these devices, an appropriate Q derived from a TEPC

may be applied to determine H at a point in adjacent tissue.

Most currently available active electronic personal dosimeters are

used routinely to measure low-LET radiation and have not been

characterized for the types and energies of the particles comprising

the fields in spacecraft. Such dosimeters, when well characterized

may perform a useful role. The ideal dosimeter would be active, store

integrated dose data and dose rate time profiles, and respond to

all field components, allowing a good determination of D·

and H·

to

adjacent tissues.

If active personal dosimeters are not used, then it may be necessary

to develop onboard readout capabilities for the passive dosimeters,

especially on long-duration ISS missions. Although onboard readout

of PNTDs is not feasible, onboard systems for readout of TLDs,

OSLDs and electronic dosimeters are certainly possible. Develop-

ment in this area is therefore recommended.

6.4 Accuracy, Performance Testing, and Calibration

6.4.1 Operational Radiation Protection Requirements on

Accuracy of Dose Measurements

6.4.1.1 Recommendations of ICRP and ICRU. International and

national requirements, where they exist, are derived, in general,


from ICRP and ICRU recommendations. ICRP (1997) states, ‘‘In

practice, it is usually possible to achieve an accuracy of about 10

percent at the 95 percent confidence level for measurements of radia-

tion fields in good laboratory conditions. In the workplace, where

the energy spectrum and orientation of the radiation field are gener-

ally not well known, the uncertainties in a measurement made with

an individual dosimeter will be significantly greater. Nonuniformity

and uncertain orientation of the radiation field will introduce errors

in the use of standard models. The overall uncertainty at the 95

percent confidence level in the estimation of effective dose around

the relevant dose limit may well be a factor of 1.5 in either direction

for photons and may be substantially greater for neutrons of uncer-

tain energy, and for electrons. Greater uncertainties are also inevita-

ble at low levels of effective dose for all qualities of radiation.’’ ICRU

(1992) recommends, ‘‘in most cases, an overall uncertainty of one

standard deviation of 30 percent should be acceptable’’ (instrumental

error only).

6.4.1.2 General Requirements. The radiation exposure of astro-

nauts is to complex, multi-component fields that are difficult to deter-

mine routinely. Nevertheless, one of the objectives of the

measurement program should be to meet the general requirement

for the total measurement uncertainty of the quantity being deter-

mined of a factor of 1.5 at the 95 percent confidence level. The total

uncertainty in a subsequent estimate of the effective dose will be

greater.

6.4.2 Tests of Instrument and Dosimeter Performance

There are two accreditation programs for routine personal dose

measurements of external radiation, namely those of NVLAP

(National Voluntary Laboratory Accreditation Program) (White

et al., 2001) that uses the American National Standards Institute

standard (ANSI, 1993) and DOELAP (U.S. Department of Energy

Laboratory Accreditation Program) (DOE, 1986). The requirements

of these programs are broadly related to the recommendations of

ICRP and ICRU. For both accreditation programs there is a limit

on the normal incidence response characteristics determined under

laboratory conditions as the sum of bias (B) and standard deviation

(S). For DOELAP the criterion is: B � S � � � 0.30, where � is

the uncertainty of the conventionally true value of the quantity being

estimated. The DOELAP test criterion is interpreted as providing

approximately 70 percent confidence that a dosimeter reading would

be within 30 percent of the conventionally true value.

6.4 ACCURACY, PERFORMANCE TESTING, AND CALIBRATION / 57

Instruments and dosimeters used to assess doses to astronauts

should meet the DOELAP or NVLAP requirements for determina-

tions of the dependence of instrument response on energy and angle

of incidence combined, plus one standard deviation, for appropriate

calibration/test radiation fields. This requirement should enable the

overall accuracy requirement of a factor of 1.5 to be met for actual

conditions of use in fields of mixed particle types with broad energy

and direction distributions.

6.4.3 Calibration

Calibration is the set of operations that establishes, under specified

conditions, the relationship between values indicated by a measure-

ment, and the corresponding known values of the quantity to be

measured (ISO, 1993). The calibration factor (N) is the factor by

which the instrument- or dosimeter-corrected reading (M) is multi-

plied to obtain the value of the quantity to be measured, determined

under reference conditions. In the case of particle fluence (&) the

calibration factor is given by N � &/M. The inverse of the calibration

factor is the response (R � M/&). The response can be determined

at other than reference conditions when R � 1/kN, where k is the

field specific correction factor. The most complete calibration proce-

dure is as part of a full type test, in which a device is exposed to a

range of radiation energies and angles of incidence in a full determi-

nation of its response characteristics, as well as to other parameters

which may influence its reading (e.g., temperature, humidity). A full

type test is extensive and time-consuming, and is therefore generally

performed on only a few samples of a particular instrument or type

of dosimeter to determine its overall properties.

For a laboratory calibration to be meaningful, it is necessary to

clearly specify the relevant calibration conditions, including the char-

acteristics of the reference radiation source, the irradiation facility,

and the conversion coefficients used. Periodic, accurate calibrations

are an essential part of any quality assurance program, as are mean-

ingful intercomparisons. General guidance on calibration procedures

is given in NCRP Report No. 112 and Report No. 127 (NCRP, 1991;

1998).

The response characteristics (ISO, 1993) of all the types of devices

should be determined prior to use. In particular, the dependence of

response on energy, angle, fluence/dose, and fluence/dose rate should

be determined for the particle types to be measured. In addition,

any interfering (influence) effects, such as those from other radiation

types, temperature, vibration or electromagnetic fields, should be


determined. This will normally be accomplished by a combination

of experiment and calculation. The response determinations should

normally be in terms of the quantity fluence. An exception would be

for the determination of photon response, for which air or tissue

kerma will be more appropriate. For the determination of the

response characteristics of personal dosimeters, some irradiations

should be performed on either an anthropomorphic phantom or a

surrogate. Sufficient data should be available to estimate depen-

dence of response on angle of incidence. Where needed and where

available, recommended fluence to D and fluence to H conversion

coefficients should be used.

The statistical uncertainty of laboratory calibrations is commonly

far less than the above uncertainties. However, the D and H response

of devices is frequently appreciably energy- and angle-dependent.

In order to assess whether the above requirements on total uncer-

tainty can be achieved in practical measurements, either determina-

tion of the response is required for the radiation field in which it is

to be used, or in a simulation of this field, or it should be possible

to calculate the response in this field from the knowledge of the field

and of the detailed energy- and angle-dependence of response of

the device.

Response data for the types of devices to be used should be deter-

mined for the following energy ranges as appropriate: protons from

10 to 800 MeV; HZE ions (e.g., helium, carbon, silicon, iron) from

50 MeV n�1 to 1 GeV n�1; electrons from 0.5 to 10 MeV; and neutrons

from 1 to 180 MeV, monoenergetic or quasi-monoenergetic, plus

response data for fields which replicate the neutron field produced

by the interactions of GCR with shielding material.

7. Role of Biodosimetry inDose Assessment

Biodosimetry is the use of a biological marker for assessing the

magnitude of an exposure to radiations or chemicals. The unique

contribution of a biodosimetry program is that it provides an individ-

ual’s dose assessment as estimated from a biological endpoint. Thus,

it includes the response to the cumulative exposure and allows for

an assessment of variations in individual sensitivity.

A number of different biological markers have been utilized for

assessing ionizing radiation exposures with the most frequent being

genetic alterations, such as chromosomal aberrations and gene

mutations, and electron spin resonance in teeth and fingernails, for

example. These methods have been employed for dose estimation

for a wide range of accidental exposures and in occupational settings,

generally with a good level of success. However, there are challenges

associated with biological dosimetry for astronauts that were not

addressed in these previous studies, specifically calibration for the

unique radiation qualities and dose distributions encountered in

space.

The choice of method depends upon the specific exposure scenarios

being estimated, namely, duration of exposure and whether it is

fractionated, together with time of sampling in relation to exposure,

number of individuals involved and speed with which results can be

obtained. In the following discussion particular attention is given to

the aspect of how readily exposure assessment could be made while

the astronauts are still in space. Several methods are discussed in

the following sections.

7.1 Electron Spin Resonance

The use of electron spin resonance with teeth or fingernails is

problematic given that there is little information on its application

with high-LET radiations and nonuniform exposures. Furthermore,

the equipment necessary for the analysis would not be available

until the return to Earth. Thus, its use would be restricted to longer-

term evaluation.

59

60 / 7. ROLE OF BIODOSIMETRY IN DOSE ASSESSMENT

Recent intercomparisons on coded samples of artificially irradiated

whole teeth indicate that with certain methodologies doses of about

100 mGy can be accurately measured (Egersdorfer et al., 1996).

However, for the same studies, a high degree of variability in accu-

racy between laboratories and for different methods was observed.

7.2 Biochemical Indicators

As noted by Muller and Streffer (1991), the use of changes in the

chemical composition of body fluids (saliva, blood, urine) for the

estimation of radiation dose is an attractive idea. Following radiation

exposure, there is a release of enzymes or degradation of proteins

and nucleic acids, leading to shifts in the concentration of a number

of substances, e.g., amylase, taurine, thymidine, deoxycytidine (Kaul

et al., 1986). The apparent attractiveness of this approach is out-

weighed by the fact that there is a high variability, within and

between individuals, of the concentration of the molecule being stud-

ied. None of the biochemical indicators that have been described

have been (or can be) used as a quantitative dosimeter. Thus, no

lower level of detection can be established for these markers.

7.3 Erythrocytes with Transferrin Receptors

A recently described approach (Gong et al., 1999) utilizes the pro-

longed life span of erythrocytes bearing transferrin receptors on their

membrane as a proposed dosimeter. The authors describe a logarith-

mic-quadratic dose response over a 1 mGy to 1.5 Gy x-ray range. At

this time the approach has only been tested in rats, and only with

x rays. A number of assay parameters and the range of interindividual

variation need to be more clearly delineated. In addition, the assay,

as described in rats, is conducted some six weeks after exposure to

allow for the prolonged life-span phenotype to develop. This assay

cannot be recommended for use on astronauts exposed to high-LET

radiations in space. This method has not been applied to humans

thus no lower limit of detection has been established.

7.4 Gene Mutation Assays

The gene mutation assay that would appear to have the greatest

potential for the assessment of exposures to ionizing radiations

7.5 CYTOGENETIC ALTERATIONS / 61

would be that of glycophorin A mutations (Bigbee et al., 1997). The

method is rapid, relatively easy to conduct, and relatively sensitive.

The drawbacks are that there is very little experience with the use

of the assay for high-LET exposures. In addition, the nonuniform

nature of the exposure would be detrimental to the use of the assay

as a whole-body dosimeter.

While the assessment of low level of detection for the glycophorin A

assay has been studied in a limited way, the study of Tucker et al.

(1997) for radiation workers at the Sellafield Nuclear Facility indi-

cated that no increase in variant frequency was observed up to

an effective dose of 1.1 Sv, with the majority of individuals having

�500 mSv, the lower detection level. In contrast, the frequency of

stable chromosomal translocations assessed by fluorescence in situ

hybridization (FISH) increased with dose over this range for the

same individuals. An additional note is that these exposures were

chronic.

7.5 Cytogenetic Alterations

7.5.1 Micronuclei

Micronuclei result from loss of whole chromosomes or acentric

chromosome fragments from daughter nuclei following cell division.

They appear as small, membrane-bounded inclusions in the cyto-

plasm. Their assessment is relatively easy and rapid. The frequency

of micronuclei has been used for estimating exposures for a number

of radiation accidents with reasonable agreement with physical esti-

mates. The drawbacks are that few studies have been conducted

with high-LET exposures, and the assay is not reliable for partial

body exposures. In addition, the sensitivity is marginal at low expo-

sure levels. Thus, the micronucleus assay would not be suitable as

a biodosimeter in astronauts exposed to space radiations.

The incorporation of the cytochalasin B technique, whereby cells

in their second post-exposure cycle can be specifically analyzed for

micronuclei, has greatly increased the sensitivity of the assay. If

individual pre-exposure levels of micronucleus frequency are avail-

able, an absorbed dose of 50 mGy can be detected; without knowledge

of this background frequency the level of detection is estimated to

be 100 mGy (Prosser et al., 1988).


7.5.2 Acentric Fragments in Prematurely Condensed

Chromosomes

Interphase chromosomes can be prematurely condensed to be

observable microscopically either by fusion with mitotic cells by

treatment with a phosphatase inhibitor (okadaic acid) and a protein

kinase (p34cdc2/cyclinB) (Prasanna et al., 2000), or using calyculin A

(Kawata et al., 2000). Structural and numerical chromosomal alter-

ations can be observed as an increase (or decrease) in centric chromo-

some number or from the presence of acentric fragments. The method

has been used in a laboratory setting to assess exposure to x and

gamma rays (Kawata et al., 2000; Prasanna et al., 2000) and several

high-LET radiations (Kawata et al., 2000). The advantages are that

nondividing cells can be used for assay and the analysis is quite

straightforward. The current drawback is that the assay has had no

application as a true biodosimeter, and issues such as responsiveness

to partial body exposures or chronic exposures have not been

addressed. Further investigation appears warranted for assessing

the value of prematurely condensed chromosomes (PCC) as a biodosi-

meter in the space radiation environment.

The fact that this PCC method has not been used in biodosimetry

means that a lower detection level has not been established. A rea-

sonable assumption is that it will be no more sensitive than other

cytogenetic methods, given similar cell numbers analyzed. The

advantage is that by analyzing interphase cells a larger number of

cells are available for analysis.

7.5.3 Chromosomal Aberrations

The method that has been used the most extensively for radiation

biodosimetry is that of the analysis of structural chromosome alter-

ations. The most recent incorporation of FISH techniques has

allowed for the assessment of symmetrical (transmissible) transloca-

tions. This allows for fairly reliable dose estimation at long times

after exposure. Chromosome aberration dosimetry has been success-

fully applied to radiation accident victims, atomic-bomb survivors,

and a wide range of occupationally and medically exposed persons.

These scenarios include chronic exposures and partial body expo-

sures for which specific methods have been developed to aid in expo-

sure estimation. The distribution of aberrations among cells is useful

in this regard. Even with this strong record of success, the application

of the method for dosimetry in astronauts in space has significant

drawbacks as evidenced by published studies (reviewed in Testard

7.7 CURRENT NASA (JSC) METHODS / 63

and Sabatier, 1999). The major problems are interindividual varia-

tions in response and the nature of the exposures by which only a

small number of cells will actually be hit. This latter concern is

accentuated when the consequence of a hit by a high-energy particle

is taken into account; such a cell might well be so damaged that it

will not contribute to the analyzed cell population.

Additional laboratory investigation is required in order to consider

using cytogenetic analysis as a reliable dosimeter for space flights.

The problem will remain that analysis is technically demanding and

time consuming.

The lower level of dose detection using chromosomal aberrations

is dependent upon the number of cells analyzed. This point is

addressed by Bauchinger (1995) who estimates that for 5,000 cells

analyzed (using a generalized background frequency for dicentrics)

a significant increase in dicentrics should be observed at about

100 mGy for a group of individuals. Twenty thousand cells would

need to be analyzed to detect 50 mGy. Knowledge of the pre-exposure

dicentric frequency can lower this detection level without increasing

the number of cells analyzed.

The study by Tucker et al. (1997) reports a lower detection level of

about 500 mSv effective dose for occupationally exposed individuals

using FISH analysis of stable aberrations. The exposure was chronic

in this case; a situation for which this type of translocation anal-

ysis is particularly applicable. Again, the sensitivity of the assay

would be increased by having available the pre-exposure aberrations

frequency.

7.6 Summary for Methods

Based upon the specific requirements for a biological dosimeter

for assessing radiation exposure to astronauts in space, namely high-

LET exposures, highly nonuniform exposure, ease of analysis, and

rapidity of obtaining reliable dose estimates, it is proposed that a

version of the premature chromosome condensation method or FISH

analysis of metaphase cells are the most likely to be viable. Addi-

tional laboratory research is needed for both to validate their utility

for the stated purpose.

7.7 Current NASA (Lyndon B. Johnson Space Center)

Methods

Until fairly recently, JSC used the conventional cytogenetic

method for biodosimetry for astronauts. The approach, still used


by many surveillance laboratories, is to assess the frequencies of

chromosomal dicentrics in Giemsa-stained metaphase preparations

and estimate dose using standard dose-response curves developed

from in vitro exposures of human lymphocytes. Most of the published

literature on estimated doses for astronauts was obtained utilizing

this method. Recently, JSC has begun to use the assessment of

chromosomal translocations in metaphase by FISH for estimating

radiation exposures to astronauts from extended periods in space

(Yang et al., 1997). The FISH method used involves chromosome

painting with probes for pairwise combinations of chromosomes 1,

2 and 4 or simultaneous use of all three probes. The use of all three

probes allows for assessment of about 24 percent of the genome. As

with the conventional method, dose estimation is achieved by utiliz-

ing standard translocation calibration curves from in vitro exposure

of human lymphocytes. The difficulty of this method is how to handle

mixed radiation exposures. The current approach is to use the x- or

gamma-ray combined calibration curve (or gamma-ray curve) that

uses preflight blood samples from all astronauts on a particular

mission taken three months or less prior to flight. The postflight

samples are typically obtained within two weeks of return. The blood

samples are grown in culture for 48 h after mitogenic stimulation,

at which time metaphase cell preparations are produced.

The sensitivity of this approach has been tested only to a limited

extent. As an example, the analysis of blood samples from six crew-

members gave in-flight estimates of 0, 90, 180, 210, 210 or 270 mGy

of equivalent gamma-ray absorbed dose to the whole body. In general,

based upon the usual number of cells analyzed (about 4,000) the

sensitivity would allow for the detection of 50 mGy and higher. The

analysis of more cells would clearly increase the sensitivity. It is also

proposed by JSC to use gamma-ray calibration curves obtained for

each individual for use in postflight dose estimation for that particular

individual. This approach takes into account individual variations

in radiation sensitivity, and the fact that background translocation

frequencies vary quite considerably among individuals.

7.8 Recommendations and Future Considerations

Use of the FISH method as described in Section 7.7 is the most

appropriate biodosimetry approach for astronauts based upon avail-

able knowledge, technical availability, and experience. JSC is collect-

ing PCC preparations with the aim of determining if their use might

be advantageous, particularly to alleviate the need for obtaining

7.8 RECOMMENDATIONS AND FUTURE CONSIDERATIONS / 65

metaphase preparations using in vitro culture (see Kawata et al.,

2000 for description of methods).

NASA should continue to use biodosimetry as an ancillary compo-

nent of radiation dose assessment for astronauts during extended

space flights. The current JSC approach of using FISH for analyzing

stable chromosomal translocations in peripheral lymphocytes both

in pre- and postflight samples appears to be providing useful informa-

tion on exposures. The establishment of calibration curves from

individual preflight blood samples increases the sensitivity. Incorpo-

rating analysis of PCC will provide larger analyzable cell samples.

Improvements that can be envisaged are using chromosome painting

probes and computer analysis that allow for assessment of transloca-

tions in all chromosomes at the same time. This method has been

used successfully for tumor analysis. Automating the various FISH

methods will increase throughput enormously.

The area of genomics or molecular profiling is receiving consider-

able attention as an approach for assessing perturbations in gene

or protein expression at the total cell level. Ongoing studies in a

number of laboratories are beginning to describe alterations in gene

expression patterns after radiation exposures over a broad dose

range (e.g., Amundson et al., 2000). Specific gene expression patterns

could be used as biodosimeters once genes have been identified whose

expression remains elevated or decreased for extended periods of

time. An automated analysis of specifically constructed gene arrays

is quite feasible.

The rapid enhancement of technologies for measuring changes

in cellular markers (in response to radiation) will mean that new

biodosimeters will be developed that will be anticipated to detect

very specific exposures, i.e., cellular fingerprints.

8. RecommendedManagement of AstronautRadiation Safety Program

While a team of experts at the Mission Control Center directs

radiation protection actions for astronauts by means of flight rules

(Section 3.1), the overall concept for a radiation protection program

for astronauts is essentially the same as for other radiation workers.

Dose management and ALARA program recommendations estab-

lished for other types of radiation workers can provide NASA man-

agement useful guidance in setting up an effective operational

radiation safety program (NCRP, 1990; 1994; 1998; 1999). Specific

recommendations for the program are provided in this Section.

8.1 Components of a Low-Earth Orbit Operational

Radiation Safety Program

The main components of an operational radiation safety program

designed to implement the principles of dose limitation and ALARA

for astronauts working in LEO are:



that significantly affect the level of radiation exposure to the

astronauts, and radiation protection decisions that significantly

influence conduct of the mission;

● to collect and record data to assess astronaut doses for individual

mission and cumulative career records; and

● to identify, plan and carry out practical ALARA actions to avoid

unnecessary levels of radiation exposure.

NASA management should implement and maintain an effective

radiation safety program with the following features: clear definition

of the goals of the program, statement of the organization’s commit-

ment to the application of the ALARA principle, statement of manage-

ment’s commitment to provide adequate budgetary support for the

program, and periodic review of the overall program effectiveness.

66

8.2 RADIATION PROTECTION PRINCIPLES / 67

The radiation safety program organizational structure and lines of

authority should be clearly delineated and easily understood by all

individuals having radiation safety responsibilities in the program.

In particular, the radiation safety roles and responsibilities of the

flight director, flight surgeon, RHO, and SRAG, for both individual

flights and for the overall program, should be clearly delineated.

This Report provides several recommendations for improvements

in monitoring and controlling radiation dose to astronauts. With an

effective operational radiation safety program, individuals within

NASA will be able to identify the individual or group to whom they

could go to ensure that implementation of a proposed ALARA action

would be considered. Responsibility should be clearly assigned for

ensuring the translation of radiation protection strategies and

instrumentation from design and development through engineering,

preparation for flight, and in-orbit use.

8.2 Radiation Protection Principles Applied to Low-Earth

Orbit Missions

As with all radiation workers, the following principles of radiation

protection apply to astronauts during LEO missions (NCRP, 1993;

2000a):

● any activity that involves radiation exposure must be justified

on the basis that the expected benefits to society exceed the

overall societal cost (justification);

● the total societal detriment from such justifiable activities or

practices is to be maintained ALARA, economic and social factors

being taken into account; and

● individual dose limits are applied to ensure that the principles

of justification and ALARA are not applied in a manner that

would result in individuals or groups of individuals exceeding

levels of acceptable risk (limitation).

Radiation exposure from space flight is justified by the fact that

it is an environmental reality associated with the exploration of

space. The dose limits established for astronauts (NCRP, 2000a)

reflect the importance given space explorations and the limited and

closely monitored population of astronauts. These two principles of

radiation protection (the first and third items above) are already

factored into the overall protection efforts provided for the astronauts

who are subject to multiple high-risk activities.

68 / 8. RECOMMENDED MANAGEMENT OF SAFETY PROGRAM

The sources of exposure that should be included in NCRP dose

limits for astronauts and the actions that constitute an appropriate

approach to immediate dose management and ALARA during LEO

missions are the subject of this Section. In this Report, ‘‘immediate

dose management’’ refers to actions taken to address the potential

for high doses that result from transient events in the space radiation

environment that could impact the conduct or completion of the

mission or mission tasks, and ALARA refers to actions taken to keep

all doses as low as reasonably achievable by balancing the mission

objectives with practical dose reduction steps.

8.3 Sources of Exposure Included in NCRP Dose Limits

for Astronauts

The dose limits for astronauts include the cumulative dose from

space flight, the dose associated with mission-related aviation activi-

ties (excluding commercial flights), the dose from biomedical

research conducted as part of the astronaut’s mission duties, and

any other occupational doses including any received prior to work

as an astronaut. The dose limits do not include normal background

radiation on Earth or radiation dose received from diagnostic and

therapeutic medical procedures conducted as part of the astronaut’s

overall health care.

Astronauts perform significant noncommercial aviation activities

as part of their training for space missions, and the doses from

such activities should be included in the cumulative career dose for

stochastic effects. The doses can be estimated based on total flight

time and average dose rate for the given activity, unless specific

measured data are available.

Dose from mission-related biomedical research prior to, during or

after a given mission should be included in the cumulative career

dose for stochastic effects because the dose is a consequence of partici-

pation in that mission. Radiation doses from diagnostic and therapeu-

tic medical procedures should not be included in the career dose

because they have personal benefit in assessing or improving the

astronaut’s health. In addition, previous medical radiation doses, from

diagnostic or therapeutic medical procedures, are assumed to have

provided the individual a greater benefit than the risk associated

with the doses and should not be used in determining qualification

for future occupational exposure. In some cases, it may be difficult to

distinguish biomedical research from standard medical care for an

astronaut. Determination of which activities are biomedical research


and which are not is made by the Institutional Review Board govern-

ing astronaut activities.

8.4 Immediate Dose Management and Basic ‘‘As Low As

Reasonably Achievable’’ Concepts

While there are many differences in the way astronauts receive

their radiation exposure from that of a typical radiation worker, the

basic concepts of radiation protection remain the same. Radiation

protection actions based on time and shielding may be utilized in

unique ways to maintain radiation exposure below the administra-

tive levels (Section 3.1) and the dose limits, and these actions also can

be used effectively in maintaining the astronaut’s radiation exposure

ALARA. Specific in-flight radiation protection actions must be based

on real-time information of space radiation environments and on

effective communication of this information to the flight control team.

Good decisions on radiation protection actions require training and

understanding of radiation fields by astronauts, by the entire flight

control team, and by the entire flight management and design team.

As in conventional radiation protection programs, decisions to take

action to avoid radiation exposure of an astronaut are easily identi-

fied when they are related directly to preventing a person from

reaching an administrative level or a dose limit. The recognition and

decision to take specific action to maintain radiation doses ALARA

are based on the organization’s commitment to continually assess

reasonable steps that can be taken to reduce doses. Thus, established

radiation protection and ALARA program models, philosophies and

lessons learned can be used to further enhance radiation safety

programs for astronauts.

Preplanning (i.e., engineering design and mission planning) and

the modification of activities in-flight are important components of

the radiation safety program. Occasionally, it may be necessary to

modify astronaut activities quickly to keep doses below administra-

tive levels or dose limits (immediate dose management). The ability

to predict the space radiation environs and to measure space radia-

tion levels in real-time are paramount to implementing immediate

dose management and ALARA actions.

Immediate dose management and ALARA actions may differ

between Space Shuttle missions and for ISS missions. In spite of

these differences, actions to reduce radiation exposure comprise a

continuum of effort to keep astronaut doses below dose limits, admin-

istrative levels, or ALARA. Longer ISS missions may result in a


considerably higher dose to the astronaut. However, ISS mission

objectives may allow more flexibility in the timing of certain proce-

dures, like an EVA, to avoid a high-dose environment.

8.4.1 Immediate Dose Management Issues

To be effective, immediate dose management in the NASA team

environment requires all individuals involved to understand the

resources available and their individual responsibilities with respect

to corresponding actions, and to establish effective lines of communi-

cation. Implementation of immediate dose management actions is

the responsibility of all team members involved with work impacting

the astronaut’s exposure to radiation. As soon as it is recognized that

a specific activity, e.g., an extended EVA, may cause an astronaut to

exceed the administrative level or the dose limit, action should be

considered to amend or terminate the activity. A written plan, nota-

bly the flight rules mechanism (Section 3.1), should contain the

implementing procedures.

8.4.2 Basic ‘‘As Low As Reasonably Achievable’’ Concepts

An effective ALARA program depends on everyone involved in

design and management of spacecraft and missions understanding

the space radiation environment and its impact on astronaut radia-

tion exposure.

The RHO should assess the opportunities to apply ALARA. All

levels of management should be involved in establishing the objec-

tives of the program to stimulate interest and involvement and to

foster ‘‘ownership’’ in activities. Real-time data on the radiation envi-

ronment during a flight are essential for the management team.

Postflight review of exposure data and assessment of actions taken

are necessary to provide the management team with feedback on

the effectiveness of their actions. The ALARA program should be

supported by senior management and should be reviewed periodi-

cally to determine its effectiveness.

8.4.3 Considerations for Spacecraft and Space Suit Design

The need for immediate dose management and ALARA actions

should be actively considered when designing or modifying the Space

Shuttle, ISS, and EVA suits.


The placement of radiation instruments should be chosen to pro-

vide the best real-time information on astronaut radiation exposure

(Section 6). Immediate dose management includes providing areas

where astronauts could be moved (i.e., to a safe haven providing

additional shielding, and/or by repositioning of the Space Shuttle),

and in the case of extreme exposures providing for emergency return

of the astronaut. To minimize career doses, optimized shielding

should be provided for those areas of higher occupancy, such as

areas used during off-duty hours and sleeping quarters. For example,

model calculations and ground-based testing indicate that hydroge-

nous compounds are most effective at attenuating GCR radiation.

Polyethylene (CH2) is lightweight, safe and simple to install, and has

recently been added to some of the sleeping quarters onboard ISS.10

8.4.4 Considerations for Preflight Planning

Once again, the historical information on space radiation environ-

ments and information on projected radiation doses are necessary

in assessing the mission plan and its impact on astronaut exposures.

Choice of the flight path has significant impact on both immediate

dose management and ALARA. Some choice may be available in the

altitude, attitude, time and duration of the flight to affect exposures.

In planning EVAs, choice may be available in the timing and dura-

tion. Full utilization of the design features should be incorporated

into the in-flight procedures at the preflight planning stage.

8.4.5 Considerations for Continuous In-Flight Review

During flights, changes can occur in the mission plan, as well as

in the space radiation environment, which can affect an astronaut’s

exposure or potential exposure. As with any management of exposure

control, the level of effort to identify dose reduction strategies, the

involvement of management review and the immediacy of action

increase as higher doses are anticipated. Astronauts have important

responsibilities in their own dose management during flights. They

are responsible for wearing the assigned dosimeters, for taking the

necessary radiation measurements, and for taking the directed

actions.

10 Cucinotta, F.A. (2001). Personal communication (Lyndon B. Johnson Space Cen-

ter, National Aeronautics and Space Administration, Houston, Texas).


A model for immediate dose management and ALARA assessment

by astronauts and mission control during LEO flights may be as

follows:

● As the mission progresses, continually identify and assess rea-

sonable opportunities to reduce astronaut doses even if the

administrative levels are not being approached (ALARA action).

● If the astronaut’s dose is anticipated to exceed a pre-established

administrative level for the mission, an assessment should be

made of actions to reduce further dose or to prevent the antici-

pated dose from being received (immediate dose management

action).

● If the astronaut’s dose is anticipated to exceed the cumulative

short-term or career limit, an immediate assessment, including

consideration of ending the mission for this astronaut, should

be made to reduce additional dose or to prevent the anticipated

dose from exceeding the short-term or career limit (immediate

dose management action).

● If an astronaut has received greater than the cumulative short-

term or career limit, then the mission should end for that astro-

naut as soon as practical (immediate dose management action).

8.4.6 Considerations for Postflight Review

The dose placed in the record for an astronaut from a given mission

may not be known fully until after completion of the mission. Some of

the radiation measuring devices used are passive devices, requiring

processing upon return to Earth. Some data gathered may not be

sent back to Earth during the mission, but stored electronically.

Upon return to Earth, these data need to be processed to provide

the final dosimetry record.

The final dose should be determined as soon as possible following

the mission and that information reviewed with the astronaut. These

actions should be completed before the astronaut’s next mission or

within three months, whichever is sooner.

A review should be made of the immediate dose management and

ALARA actions associated with the mission. The review should be

taken into account when determining the need for changes in design

criteria, preflight planning, and flight rules used for future flights.

8.5 Radiation Safety Training for NASA Personnel

An NCRP symposium (NCRP, 1997) identified ‘‘a universal need

to educate those involved in applying radiation dose limits to ensure

8.5 RADIATION SAFETY TRAINING / 73

that they understand that the limits are not there to ‘be used up’;

rather their purpose is to serve as a constraint on operation and as

a guide for the design of the associated radiation protection systems.’’

All NASA personnel whose work has impact on astronaut exposure to

radiation should be trained in the techniques of radiation protection,

with emphasis on implementation of immediate dose management

and ALARA. The scope and depth of this training should be related

to the corresponding level of impact the individual may have on

astronaut dose.

8.5.1 Astronauts

Astronaut training concerning radiation is described in NASA

(2000c) for Space Shuttle crew and NASA (2000d) for ISS crew.

Astronauts currently receive radiation training at various times

throughout their career. During the astronaut candidate year they

receive a 1 h overview of the JSC Space Flight Radiation Program.

When training for a specific flight, Space Shuttle astronauts are

instructed on use of the radiation monitoring equipment, and in the

course of this discussion, they receive a refresher on the radiation

environment. ISS crewmembers receive the same training as Space

Shuttle crewmembers, but also receive more detailed information

on their projected dose, and the implications and possible health

consequences of this exposure.

Ten days before the flight, during the medical exam, crewmembers

are advised of the projected exposures and associated risks of both

in cabin and EVA exposure. They are also reminded to wear their

personal dosimeters throughout the mission. Space Shuttle crew-

members receive approximately 30 min of radiation training in the

preflight period. ISS crewmembers receive 2.5 h of radiation training

prior to flight. The dose received is communicated to the astronaut

during a postflight debrief.

The astronaut’s radiation protection training should focus on the

following:

● biological risks of radiation exposure, particularly for the high-

LET space radiations

● measurement of radiation and dose

● operation of and proper wearing of personal radiation measur-

ing devices

● techniques an astronaut can use to reduce dose in-flight

● specialized training specific to a given mission, such as operation

and maintenance of an area radiation monitor


● procedures and administrative controls including dose limits

and the flight rules (immediate dose management and ALARA)

● terminology associated with dose management activities to

ensure effective communications

● NASA’s operational radiation safety program and how ALARA

actions are proposed and implemented

● review of actual events, especially those involving high tran-

sient doses

8.5.2 Flight Directors

Flight director training concerning radiation is described in the

Flight Director Certification Guide (NASA, 1998). The flight director

is the individual responsible for a given mission. The current radia-

tion safety training for flight directors consists of an introductory

briefing at the start of their overall training, which includes radiation

protection terminology, description of the space radiation, radiation

risk versus dose, dose limits, and the ALARA philosophy. Beyond

this briefing, they generally receive little formal training in radiation

issues. Occasionally a special briefing is provided when there are

significant changes or significant updates to the flight rules. Most

training is informal, but can be substantial, occurring during devel-

opment of radiation flight rules and in the course of mission planning,

mission simulation, and flight training.

Like managers and supervisors of other radiation workers, the

flight directors should receive radiation protection training that par-

allels those they supervise, but not necessarily to the same depth.

The flight director’s radiation protection training should focus on

the following areas:

● policies and procedures affecting astronaut’s radiation dose

● implementation of immediate dose management and ALARA

actions and their importance

● interfaces of different NASA groups so that they are able to

identify and resolve conflicts in coordinating radiation protec-

tion activities





● review of actual events, especially those involving high transient

doses and those involving the actions of a flight director

8.5 RADIATION SAFETY TRAINING / 75

8.5.3 Flight Surgeons

Flight surgeon training concerning radiation is described in the

ISS Crew Surgeon Training Certification Plan (NASA, 1999).

A flight surgeon is assigned to each mission and is responsible for

that crew’s health and safety. The current radiation safety training

for flight surgeons consists of an initial 3 to 4 h of training in radia-

tion issues, including the ALARA principle. The RHO provides quar-

terly briefings to the flight surgeons as refresher training and to keep

them informed of changes and new radiation issues and concerns. In

the course of their duties flight surgeons may receive substantial

informal training as part of the team that is responsible for crew

health and safety. They also receive informal training through

involvement in mission planning and as participants in a variety of

activities that address radiation issues.

As with the flight directors, the flight surgeon should receive radia-

tion protection training that parallels those of the astronauts, and

particularly in regard to health effects. The flight surgeon’s radiation

protection training should focus on the following areas:

● biological risks of radiation exposure, particularly for the high-

LET space radiations

● procedures and administrative controls including dose limits

and flight rules (immediate dose management and ALARA)

● implementation of immediate dose management and ALARA

actions, and their importance

● interfaces of different NASA groups so that they are able to

identify and resolve conflicts regarding health consequences of

ongoing or projected radiation doses






doses and those involving the actions of a flight surgeon

8.5.4 Radiation Health Officer and Radiation Safety

Support Groups

The RHO and individuals in the radiation safety support groups

should have training equivalent to radiation protection professionals

in other professions who are assigned equivalent types of radiation

safety duties (NCRP, 1998; 2000b). Radiation protection training must

also be specific to the special circumstances of the space radiation


environment. They should also have training on the specifics of

NASA’s operational radiation safety program for LEO.

8.5.5 Other Supporting Specialists

Other individuals whose work can affect the astronaut’s radiation

exposure should have radiation safety training that is commensurate

with the level of effect their work might have on astronaut dose.

Among these individuals are engineers, designers and mission plan-

ners. These individuals’ radiation protection training should focus

on developing the following knowledge:

● how their work can affect astronaut dose

● what the immediate dose management and ALARA concepts

are, and how they can help implement them




doses and those involving the actions of individuals with similar

work responsibilities

9. Radiation Safety Records

The principal reason for keeping radiation safety records is protec-

tion of the astronauts. This Section describes the elements of a pro-

gram for operational radiation safety records, but does not prescribe

a particular methodology for construction or maintenance of the

records.

In this Report the terms ‘‘dosimetry records’’ and ‘‘radiation safety

records’’ are used. Dosimetry records are the collections of informa-

tion that document the radiation doses that astronauts receive from

space flight, mission-related aviation activities, mission-related bio-

medical research, and previous occupational work. The dosimetry

records contain, or are linked to, all the basic information that is

necessary to obtain the radiation doses, e.g., the personal dosimetry

measurement package (Section 6.3). Radiation safety records are the

broader collections of information that relate to radiation exposure of

the astronauts. The radiation safety records consist of multiple

sources of information that are linked (i.e., the dosimetry records

and their supporting data and files, and other information that bears

on the radiation exposure of astronauts). When the term ‘‘records’’

is used, the meaning is general.

The objectives of the overall radiation safety record-keeping pro-

gram are to:

● aid in the protection of astronauts

● evaluate the effectiveness of radiation protection programs

● advance the collection of data that are accurate, reliable, confi-

dential, and retrievable

● provide a uniform set of records

Dosimetry records for each astronaut should be kept in a manner

to satisfy the following purposes:

● inform astronauts of their mission and cumulative radiation

doses

● make decisions on crew selection and mission planning

● provide information for making decisions about the flight plan

● evaluate the operational radiation safety program to ensure

effective program operation

● demonstrate compliance with administrative levels and dose

limits

77

78 / 9. RADIATION SAFETY RECORDS

● provide reliable data for epidemiological studies

● provide information for making or contesting claims for radia-

tion-induced injury

All records should be maintained in accordance with the highest

standards of record keeping (ANSI, 1999; Fienberg et al., 1985;

NCRP, 1992) and should be legible, accurate, reliable and interpret-

able. Individuals who maintain the records should be diligent in

protecting the identity of astronauts and any proprietary informa-

tion. Information that may be separate from the dosimetry records,

but which is important to interpret doses or recalculate dose quanti-

ties, should be linked in a manner that assures all information will

be available for future interpretations or recalculations.

Records may take the form of hard copy, microform, or computer

readable media. Each form has its own strengths and limitations

and may be used for different purposes. Of particular importance is

the conversion of old data to new formats when changes in computer

hardware render old formats unreadable. Records should be stored

in a durable enough manner to limit the risk of loss from theft, fire

and water damage. A listing of content and location of records should

be maintained to facilitate their retrieval.

When data are stored on computer readable media, the original

or reproduced records from which the computer records were derived

should be maintained separately to help interpret ambiguous com-

puter data and to verify computer records if they become damaged

or unreadable. To enhance reliability, computer records should be

backed up on a regular basis.

9.1 Content of Records

Standard operating procedures should be written to enable those

responsible for generation and maintenance of the records to docu-

ment and retrieve information in a timely and reliable manner.

Procedures should address the indexing, linkage and storage of docu-

ments. Standard codes and coding conventions should be used, e.g.,

North American Industry Classification System (OMB, 1997) and

Standard Occupational Classifications Manual (OMB, 2000).

Information should be documented in as specific or quantitative

a form as is available. However, qualitative information should also

be kept when it may be useful for interpretation of the records or

recalculation of radiation doses. When quantities are derived or cal-

culated from measured quantities, both the measurement data and

the quantities derived from them, along with references to the

9.1 CONTENT OF RECORDS / 79

method used in the derivation should be maintained along with all

information that may be useful in the interpretation of the measure-

ment data. Missing information should be addressed in the records,

and data that were estimated or calculated rather than measured

directly should be documented. The records should contain adequate

information to facilitate the interpretation of an empty field or zero

value. Computations and estimates should be supported by records

that provide the method used to perform the calculation or to make

the estimate.

Dosimetry records and associated reports should include calendar

dates of radiation exposures and starting and stopping dates for

radiation monitoring as well as the date the records were generated.

These records should include the magnitude and dates of radiation

exposure and linkage to other records that identify the sources of

exposure, place where the exposure occurred, description of radiation

fields, description of physical factors that may alter the field, and

information about other potentially harmful exposures.

The dosimetry records should contain sufficient information to per-

mit unique identification of the individuals involved and information

on job or task assignments that should be updated at least annually.

Information on general work location (e.g., Space Shuttle, ISS, EVA)

should be linked to the dosimetry records to facilitate correlation

analysis. Other information, such as non-mission related diagnostic

and therapeutic medical radiation exposures, should be linked to the

dosimetry records, but the radiation doses from the medical proce-

dures should not be added to occupational doses either for planning

purposes or to limit occupational exposure. Medical radiation expo-

sures should be maintained in the medical department.

Information important to standard methodology, collection of data,

or calculation of radiation doses may exist in other files that are

separate from the dosimetry records. These files should be linked to

the dosimetry records in a manner that facilitates re-evaluation and

auditing of astronaut radiation doses. Files to be kept in or linked

to the dosimetry records include:

● standard operating procedures on how to maintain the record

● standard operating procedures on how to compute the radiation

dose of record

● personal data for each astronaut, e.g., age, sex, height and weight

● flight plans

● steps taken to avoid or minimize radiation exposure during a

flight, e.g., changes in the flight plan such as reduction of time

spent outside the spacecraft or moving to a shielded portion of

the spacecraft during increased solar activity


● raw data on the radiation environment and spacecraft character-

istics and orientation, which were collected at the time of the

astronaut’s exposure to the radiation

● information on radiation detectors and measurement equipment

including type, serial number, and calibration data

● biodosimetry data and computations that utilize these data

● computations of radiation doses, including method used to

account for radiation quality and the tissue weighting factor (wT)

● reports of monthly, annual and career radiation doses from space

flights for each astronaut

● radiation dose estimates for biomedical activities that are associ-

ated with specific flights or research protocols

● radiation dose estimates from associated noncommercial avia-

tion activities undertaken as part of mission requirements

9.1.1 Records of Career Doses

Dosimetry records constitute the formal documentation for each

astronaut’s space-related radiation exposure history and should con-

tain the cumulative dose from space flight, mission-related aviation

activities, and mission-related biomedical research. Since the radia-

tion doses for each astronaut may not be assigned until well after

completion of a flight (or may be revised as new information becomes

available), NCRP recommends that the dosimetry records be updated

as soon as possible after the radiation doses are assigned (or

changed). The new doses should be communicated to the astronauts.

The dosimetry records should contain, or be linked to, all the

basic information that is necessary to obtain the required dose-limit

quantities (GT and E). In addition to the astronaut’s measured radia-

tion doses (e.g., from personal dosimeters or area monitors) for each

point in time or monitoring period, the dosimetry records should

contain the assigned value of DT for each tissue or organ and the

method by which HT for each organ or tissue was obtained for use

in the computation of E for that astronaut. The computed value of

E should be linked to the standard operating procedure that

describes the computation. The value of GT for skin, lens of the eye,

and bone marrow should be based on surface measurements, which

are adjusted as described in Section 5. The dosimetry records should

also be linked to information on radiation exposure conditions at the

location and time of exposure.

Normally, astronauts are not exposed to radioactive materials that

may be ingested, inhaled or absorbed through the skin. Participation

in biomedical research studies that use radionuclides as tracers

9.1 CONTENT OF RECORDS / 81

would be an exception to this. When astronauts are monitored for

internal radiation sources, the dosimetry records should contain the

basic physical measurements (e.g., activity) used to determine DT

to organs and tissues. Information on mode of intake and, where

applicable, data on particle size distribution, chemical form, and

chelation or decorporation therapy, as well as, procedures, bioassay

data, biokinetic and dosimetric models, and other parameters used

to determine DT should be linked to the dosimetry records. For inter-

nally deposited radionuclides from these biomedical studies, HT

should be obtained using NCRP (1993) radiation weighting factors

(wR) and then used to compute E.

If an astronaut received occupational exposure from previous

employment or military service, information on the identity of each

former employer and the radiation exposure for each period of

employment should be obtained. This information may be included

directly in each astronaut’s dosimetry records or may be kept in

other files that are linked to the astronaut’s dosimetry records.

Annually, each astronaut should receive a confidential report of

his or her radiation dose assessment and information on the signifi-

cance of the doses. The report should include career radiation doses

in terms of E for stochastic effects and monthly and annual doses

in terms of GT for deterministic effects to skin, lens of the eye, and

bone marrow.

In addition to occupational radiation dose, the report may contain

information on radiation doses from diagnostic or therapeutic medi-

cal procedures. Astronauts should be provided with information that

describes the significance of these nonoccupational doses and the

potential biological effects.

9.1.2 Records of Prospective and Retrospective Studies

The probability that particular records will need to be retrieved

depends on the purpose for which the records are to be used. The

four major potential uses of archived dosimetry records are:

● evaluation of the radiation safety program

● compliance with the dose limits

● use in epidemiological research

● use in litigation

If dosimetry records for short term or career doses are updated

(revised) in a timely fashion whenever relevant new information

(e.g., new values for wT) becomes available, they will have greater

value for prospective use in planning astronaut activities and in


retrospective studies. Linkages to other information as described

above (Section 9.1) will facilitate broader analysis of calculations,

computations, flight control plans, and other variables that may

have affected the recorded doses.

9.2 Continuity of Records Over Time

Dosimetry records should provide continuity of career radiation

doses from past, present and future missions. Continuity is impor-

tant to evaluate trends in radiation doses among astronauts over

time, for prospective estimation of doses, and for assessment of cumu-

lative career doses. As discussed in Section 6.3.3, NCRP recommends

that the dosimetry records include the low- and high-LET compo-

nents of the radiation field. If possible, the dosimetry records for the

high-LET component should be divided into its neutron plus high-

energy secondary particle (from protons) component and its HZE-

particle component.

9.2.1 Existing Records and Unanalyzed Data Files

As a general recommendation, NCRP advises that any existing

data related to space missions be examined and compared to dosime-

try records to confirm that radiation dose estimates are based on

all available data. If previously unanalyzed data are found, NCRP

recommends that they be analyzed and that a determination be

made of the value of the data in estimating astronaut doses.

The amount of data that may exist, which could change the current

career dose estimate for astronauts, may be considerable. Analysis

of all these data may or may not have a significant effect on the

recorded annual dose estimate. Therefore, NCRP recommends that

the data be identified and that representative samples of the data

be fully analyzed to determine the extent of their effect on current

dose estimates.

9.2.2 Future Adjustment of Records

Dosimetry records should be updated retrospectively whenever

there is a systematic change in methodology (e.g., method of account-

ing for radiation qualities or wT) or new information becomes avail-

able (e.g., previously unanalyzed data). Any data that could affect

the estimation of astronaut doses should be analyzed as soon as

9.3 RETENTION OF RECORDS / 83

possible or retrospectively to provide the best estimate. NCRP recom-

mends that astronaut dose estimates be adjusted whenever differ-

ences in revised dose assessments exceed 30 percent of the original

dose assessment (see discussion of accuracy in Sections 6.4.1 and

6.4.2).

9.3 Retention of Records

Determination of the retention period depends on the conse-

quences of having or not having the information available when it

is needed. Regulations specify minimum retention periods for the

first two potential uses given in Section 9.1.2, generally within a

time frame of years or possibly the lifetime of the individual. Conse-

quences associated with potential future need for records for epide-

miological research and litigation are difficult to predict. Need of

records for these purposes may not arise for decades. NCRP (1992)

recommends that records be maintained for specific periods of time

appropriate for their intended purposes. Because of the unique fea-

tures of astronaut radiation exposures, NCRP recommends at this

time that records be maintained indefinitely.

Appendix A

Time-Dependent Variationsin the Space RadiationEnvironment and the Needfor Active Real-TimeMonitoring

A.1 Background

Temporal changes in the external radiation environment occur by

two major processes and one minor process:

● precession of the orbit within Earth’s asymmetric magnetic field

● modulation of the sources by solar activity (both short- and long-

term variations)

● change in altitude due to atmospheric drag and re-boost

In addition, there are local spatial variations within the Space

Shuttle, ISS, or space suit due to the relative differences in shielding

at locations visited by the astronaut. This requires an accurate

knowledge of the time spent by the astronaut within the spatially

distributed fields and the temporal field changes during those visits.

There are uncertainties in the external radiation environment.

The properties of GCR are well known and the transmission through

the geomagnetic field is adequately understood except during brief

periods of large geomagnetic storms. Such storms result in little

uncertainty in the accumulated fluence over a Space Shuttle or ISS

mission. The neutron albedo is directly correlated to the GCR intensity

and has been well characterized by satellite and balloon measurements

and theoretical models. The LEO trapped proton environment is rel-

atively stable with mainly long-term variations over the solar cycle

but this time variation is poorly represented in current environmental

84

A.1 BACKGROUND / 85

models and is a major source of uncertainty in Space Shuttle

or ISS exposures. In addition, the trapped radiations are highly

anisotropic; this anisotropy is not part of the usual environmental

models.

SPEs cause short-term enhancements in the extraterrestrial radi-

ation environment, the effects of which are mainly limited to the

geomagnetic polar regions. Due to orbit precession, the worst-case

SPE exposure occurs when the line of nodes (i.e., the intersection of

the equatorial plane with the orbital plane) is near 210 degrees

longitude. Only the two prior and two succeeding orbits are impor-

tant to exposure if the 210 degree line of nodes is coincident with

the peak of the SPE intensity. The time of particle arrival relative

to orbital precession is critical to exposure evaluation. A large geo-

magnetic storm could greatly enhance the radiation exposure if the

peak storm is at the SPE peak intensity while the orbit plane is

aligned in the worst-case orbit. The expected exposures from an SPE

and their associated uncertainties depend on many variables. The

above considerations apply to both Space Shuttle and ISS and during

EVA. In addition to these sources, the trapped electrons are impor-

tant for EVA. They are filtered from the interior of the Space Shuttle

or ISS radiation environment by the micrometeoroid/debris bumper

and pressure vessel and the resulting bremsstrahlung is unimpor-

tant compared to other LEO sources. The trapped electrons are

highly variable with relativistic electron precipitation events occur-

ring every few to several weeks. During these events, the fluence

can rise suddenly by up to four orders of magnitude with slow decline

over several days. These events are most important to skin exposures

during EVA. The helmet provides significant protection to the lens

of the eye especially with the visor in the down position.

Active detectors provide information, in real-time, on radiation

fields inside and outside the spacecraft. This information is used for

direct determination of D and H. Active dosimeters can be either

area monitors or personal dosimeters. Active instruments can also

measure particle fluence and fluence rates as a function of particle

charge and energy. Active devices, by virtue of their immediate

availability, are valuable sources of data needed for timely imple-

mentation of immediate dose management and routine ALARA.

The instrumentation should be capable of providing dose rate as

well as integrated doses with time resolution from several minutes

to several days.

Information should, where possible, be immediately available to

astronauts in space as well as through downlink to the ground where

more extensive analysis can be performed.

86 / APPENDIX A

Active dosimeters should be used inside and outside of the space-

craft. Some serve as area monitors at fixed locations while others

are portable area monitors that can be moved to various locations,

or even carried on the astronaut’s person, including during EVA.

A.2 Types of Active Detectors Used in Space

A.2.1 Tissue Equivalent Proportional Counters

A TEPC is a gas filled detector that is operated in the proportional

mode such that the signal generated is proportional to the initial

ionization generated in the gas (Rossi and Rosenweig, 1955). The

tissue equivalence of a TEPC is based on the composition of the wall

that defines the gas cavity as well as the composition of the gas

itself. Generally the wall is fabricated from a conducting plastic such

as A-150� (Standard Imaging Inc., Middleton, Wisconsin). Since a

low-density gas is used to simulate very small volumes of material,

the Bragg-Gray cavity theory would apply for any gas mixture. How-

ever, in order to avoid variations in stopping power ratios (Swall/Sgas)

for the vast array of incident particles, the gas is also chosen to have

an atomic composition of soft tissue as well as retain the properties

necessary for gas multiplication in a high electric field. It is usually

a combination of carbon dioxide, argon, and either methane or

propane.

Another feature of a TEPC is that the quantity of gas in the cavity

is carefully regulated. The reason for this is that energy deposition

from the passage of a charged particle in the gas can simulate energy

deposition by that particle in a volume of tissue having dimensions

similar to a mammalian cell. Thus, energy deposition events mea-

sured by TEPC should reflect the pattern of energy deposition in

microscopic volumes. This in turn provides information on the qual-

ity of the incident radiations.

The basis for simulating small volumes of tissue rests in the fact

that cross sections for ionization at the atomic and molecular level

are independent of density. For example, energy transfer is con-

served providing that:

�t xt � �g xg, (A.1)

where:

A.2 TYPES OF ACTIVE DETECTORS USED IN SPACE / 87

xt � the thickness of tissue to be simulated

�t � the density of tissue, taken to be 1 g cm�3

xg � the thickness of the gas cavity in the detector

�g � the density of the gas in the detector

A common case is when a spherical or cylindrical detector with a

diameter of 10 mm is used to simulate a volume of tissue with a

diameter of 1 �m. For this situation, the gas pressure in the cavity

would be about 6,700 Pa. These conditions can be achieved routinely

and provide signals that are reasonably easy to process and record.

They have been successfully used for dosimetry in mixed fields con-

sisting of fast neutrons and photons, as well as for manned space

missions.

For heavy charged particles passing through a very small amount

of material, the energy lost is small and LET remains unchanged.

If all of the energy transferred to the volume of material is absorbed

within the volume then:

� � L x (A.2)

where:

� � the energy deposited (keV)

L � LET (keV �m�1)

x � the path length through the material (�m)

Thus, energy deposition in a cell for this case depends on both

LET and path length or trajectory through the volume. This should

also be the case for a TEPC.

Microdosimetry was developed to accommodate these principles

(ICRU, 1983). It is known that there is an association between RBE

and LET. This has been extended to radiation protection where Q

is also a function of LET. Thus, the energy deposition in a cell should

be divided by a length if it is to be related to LET. However, the

length through the cell may not be known and this changes from

event to event. Thus, a quantity called lineal energy (y) was defined

such that:

y ��

�, (A.3)

where:

y � the lineal energy for an event (keV �m�1)

� � the energy deposition associated with the event (keV)

� � the mean chord length for the distribution of possible trajecto-

ries through the simulated volume of tissue (�m)

88 / APPENDIX A

Thus, y is a random variable that depends on the distribution of

energy deposition events in the volume of interest and a constant

that depends on the geometry of the tissue. The mean chord length

was chosen for the denominator of y because for a distribution of

random events with constant LET, the mean value of y will be numer-

ically equivalent to LET.

For the case where LET of the particles is constant, the response

of a TEPC will reflect the chord length distribution for the simulated

volume of interest. Figure A.1 shows the expected distribution of

f (y) for a uniform isotropic particle fluence [&(L)] having a single

value of L equal to 215 keV �m�1. Curve 1 represents a sphere with

a diameter of 1 �m and Curve 2 represents a cylinder with diameter

and length equal to 1 �m. Although the mean chord length happens

to be the same for both of these distributions, the pattern of energy

deposition events is distinctly different.

It is clear that even for an idealized case, the measured distribution

f (y) will not be a direct representation of the distribution of LET for

the incident &(L). The distribution of f (y) is a convolution of &(L)

and f (x), the distribution of path lengths (x). Thus, if f (x) is known,

it should be possible to deconvolute f (y) and obtain &(L). When f (x)

is the triangular distribution for a sphere, the deconvolution reduces

to an analytical expression that is easy to solve. Other distributions

of f (x), including that for a cylinder, usually require complicated

numerical procedures.

The situation for heavy nuclei is more complicated. The energy

deposition does not occur in a limited region surrounding the track.

High-energy delta rays remove energy away from the axis of the

track. Some of the energy can actually escape from the volume of

interest. For the same reason, there can be energy deposition in a

volume of material that was not traversed by a track at all. In

addition to these factors, there is additional variance in the measured

signals due to energy straggling, gas multiplication, and effects due

to the fact that the density of the wall surrounding the cavity is

40,000 times greater than the density of the gas.

Curve 3 in the following Figure A.1 is the distribution f (y) mea-

sured in a spherical TEPC exposed to 56Fe particles at 400 MeV n�1

(L � 215 keV �m�1) (Gersey et al., 2002). It clearly is not a simple

triangle. There is a broad peak in the distribution that resembles a

triangle. These events are from particles that pass through the gas

cavity, but the energy absorbed in the cavity is less than L times x

because some high-energy delta rays escape the sensitive volume.

There is also a pronounced peak of events with very small values of

energy deposition. These events correspond to particles that do not

pass through the gas cavity but intersect the plastic wall surrounding


f(y) 4

32

1

y (keV µm –1)

0 100 200 300 400 500

Fig. A.1. Distributions of y in a TEPC for a uniform fluence of 56Fe

at 400 MeV n�1. Curves 1 and 2 represent the case where energy deposited

is equal to L times the chord length in a sphere (Curve 1) and a cylinder

(Curve 2). Curve 3 is data from measurements with a spherical detector.

Curve 4 is the result from a calculation using restricted LET (L�) and

straggling.

the cavity. Some of the delta rays originating in the plastic manage

to enter the cavity and deposit energy. There is also a tail of events

that corresponds to very large energy depositions. These events are

caused by particles that just graze the inside wall of the detector. This

produces a flood of low-energy delta rays that combine to generate a

very large energy deposition event in the gas (Rademacher et al.,

1998). These patterns have been observed for 56Fe from 200 to

1,000 MeV n�1 as well as for other ions such as 28Si, 16O, and 12C.

The magnitude of these effects will change with the charge and

velocity of the heavy ion making the deconvolution process even

more complicated when L has a large range of values.

A model has been developed to compute energy deposition by HZE

particles in order to predict the measured distribution of f (y) by a

TEPC exposed to space radiation (Shinn et al., 1999; Xapsos, 1992).

This computation takes into account chord-length distributions

through the detector similar to Curves 1 and 2 in Figure A.1. Energy

deposition is obtained using a model for restricted LET (L�) to simu-

late delta-ray escape and a lognormal distribution of energy lost

along the track to simulate particle straggling. The results of this

model for a 1 �m diameter sphere are shown as Curve 4 in

Figure A.1. When compared with the expected triangle for a sphere,

there is a decrease in energy deposition as well as a smoothing effect

due to straggling. The model does not however reproduce the data

90 / APPENDIX A

corresponding to very low and very high-energy depositions gener-

ated in the wall of a TEPC.

Although TEPC does not provide a direct measurement of LET

distribution of the incident particles, it does provide valuable infor-

mation relating to the quality of the radiation, and therefore allows

an estimate of H to a small mass of tissue at the location of TEPC.

One important issue is related to the integrated spectrum of energy

deposition in terms of D. Energy that is lost by the escape of delta

rays when the particles pass through the sensitive volume must be

replaced by delta rays that originate outside of the volume (i.e., from

the wall). Table A.1 shows results for the fraction of D obtained for

a uniform fluence of 56Fe particles at 400 MeV n�1 from the data

shown in Figure A.1. Also shown are results for the first and second

moments of the lineal energy.

The measured data for 56Fe at 400 MeV n�1 indicate that TEPC does

satisfy charged-particle equilibrium and therefore gives a correct

estimate of D to tissue for this HZE particle. However the restricted

stopping power model underestimates the dose by about 20 percent

because it does not consider energy transfer from particles that pass

near but not through the sensitive volume of the detector. Additional

experiments and data will be required to determine the extent to

which charged-particle equilibrium is achieved for HZE particles

other than 56Fe.

As expected, the frequency-averaged lineal energy (yf) is numeri-

cally equivalent to LET for the idealized distributions with a sphere

and cylinder. However this relationship is not preserved for the 56Fe

data or L� model. This is further confirmation that f (y) is not a

surrogate for &(L). It is however interesting that the dose-averaged

lineal energy (yD) is also different for all four distributions but the

value from the 56Fe data is numerically equivalent to LET. This has

been demonstrated for 56Fe from 200 to 1,000 MeV n�1 (Gersey et al.,

TABLE A.1—Estimates of the fraction of D, yf and yD associated

with the curves shown in Figure A.1.

yf yD

D (%) (keV �m�1)

LET in spherea 100 215 241

LET in cylindera 100 215 270

Fe-56 data 104 146 216

L� model 81 174 193

aRefers to the ideal case where energy deposited is equal to LET times

chord length.


2002). Thus, it is possible that even though the measured spectrum

from a TEPC is not a direct determination of LET, yD is coupled to

LET in a way that is useful for radiation protection purposes.

A TEPC is also sensitive to high-energy neutrons (Badhwar et al.,

2000; in press). Charged particles generated in the walls of a TEPC

will deposit energy in the gas cavity. The pattern of energy deposition

is similar to charged particles originating outside of the detector.

Thus, a TEPC automatically includes the contribution from neutrons

that either interact in the detector or elsewhere in the spacecraft.

A TEPC could be placed within a hydrogenous phantom to create

the charged-particle intensity that would be expected from neutron

interactions in a person.

TEPC systems currently used by NASA resolve energy deposi-

tion events in the range from a fraction of a kiloelectron volt

(y � 0.3 keV �m�1) to 800 keV (y � 1,000 keV �m�1). The lower bound

is determined by electronic noise. The upper bound by saturation of

shaping amplifiers. As electronic components improve, it should be

possible to reduce noise and extend the range of these detectors.

A.2.2 Semiconductor Detectors

Detectors based on thin semiconductor crystals measure the ion-

ization energy loss of an incident particle. Energy deposited within

the detector can be collected and analyzed in real-time, and data

from one or more such detectors can be convoluted to determine a

particle’s charge and total energy. The average energy necessary to

create an electron hole pair (ion pair) is approximately 10 times

lower in a silicon diode compared with gas. This results in improved

energy resolution and thus provides an opportunity for precise esti-

mates of LET for the incident particles.

The detectors are planar with a thickness ranging from 50 to

5,000 �m and diameters ranging from 10 to 60 mm. High-energy

charged particles pass through the sensitive volume. The magnitude

of the signal (dE) registered by the detector is related to the trajectory

(angle of incidence) and LET of the incident particle. Energy resolu-

tion improves as dE increases. If the acceptance is restricted to small

angles, the path through the detector is approximately equal to the

thickness (xd) and L � dE/xd. Under this approximation, a detector

can yield D, Q and H for charged particles. If the variation in path

length is restricted to 10 percent, the incidence angle must be limited

to 26 degrees, which corresponds to a solid angle acceptance of 0.1.

This means that approximately 90 percent of the incident particles

must be rejected.

92 / APPENDIX A

Some of these restrictions can be eliminated by adding one or

more position sensitive solid-state detectors (PSD) to the system.

For example, resistive charge division is used to determine the loca-

tion of the incident particle in the plane of the detector. Data from

two or more PSDs placed a fixed distance apart can be used to

reconstruct the trajectory (incident angle) of the particle and thus

the true path length (dx) through the dE detector. The dE/dx thus

obtained is L�. There must be sufficient separation between the PSDs

in order to measure the incident angle. Limits on the sensitive area,

and separation between PSDs have the effect of restricting the angu-

lar acceptance of this system.

For dosimetry purposes alone, a thick detector (xd � 500 �m) with

an area of about 300 mm2 (diameter �20 mm) should be sufficient

for a first approximation to D and Q, and devices of this type that

are sufficiently compact to be worn on an astronaut’s person exist.

For this case the interpretation of the pulse height spectrum in terms

of H must take into consideration that some of the particles were

incident directly upon the dosimeter and some went through the

astronaut’s body. In addition, this system cannot distinguish slow

particles that stop in the detector from those that penetrate. Thus,

dE may be correct for estimating D, but setting dx � xd may under-

estimate L.

Detector systems that contain several solid-state detectors can

eliminate some of the problems mentioned above (Badhwar et al.,

1995b; Doke et al., 2001). These arrangements are often called parti-

cle telescopes since they view a selected region of space. PSDs can

be used to define the trajectory. The distribution of energy loss in

successive detectors is a unique signature of a particle’s charge and

energy, and instruments of this type have been flown on satellites;

however, the number of detectors required may render this method

impractical for dosimetry. Combining a thin detector with one suffi-

ciently thick to stop a charged particle is effective for charge identifi-

cation, but also may be impractical for many of the highly energetic

particles found in space.

There have been attempts to make omnidirectional detector

arrays, in contrast to the collinear pattern used in particle telescopes.

One design is based on making a cube from silicon strip detectors

(Yoshihira et al., 2000). The detector would measure dE and position

of the incident particle as it entered and exited the cube. The position

information is used to compute the trajectory of the particle and dx

in each detector. This could then provide &(L)/d� as a function of

direction or &(L) integrated over all directions.

A single active silicon detector system has been made compact

enough to serve as a personal dosimeter (Badhwar, 2000). It provides


on-demand readout of accumulated D as well as D·. Power is obtained

from a stack of lithium batteries that last for about 28 d. It can be

modified to operate from space suit power during an EVA.

A.2.3 Cerenkov Counters

An alternative approach is to make a determination of dE/dx and

velocity (�). One way to do this is by measuring Cerenkov radiation.

The procedure is based on using thin solid-state detectors to measure

dE/dx and a Cerenkov counter to measure the photon emission

(dN/dx) in a gas or liquid. These are related to the incident particle

in the following manner:

dE/dx �Z2

�2, (A.4)

dN/dx � Z2 1 �1

�2n2, (A.5)

where n is the refractive index of the medium in the Cerenkov

counter. The two equations can be solved simultaneously to obtain

the charge of the incident particle (Z) and velocity (�).

The Cerenkov counters used in these applications generally have

a threshold of about � � 0.5 (200 MeV n�1) and saturate at about

� � 0.75 (500 MeV n�1).

A.2.4 Ionization Chambers

Ionization chambers can be used to measure D. They, however,

do not give information on the quality of the incident radiation for

mixed radiation fields. There is also a problem of recombination

along the track of high-LET particles.

However, ionization chambers can be used as screening detectors

for solar events or to monitor enhanced electron intensity following

magnetic storms. They could be employed as area monitors inside

and outside of the spacecraft in order to confirm that dose rates do

not approach or exceed specified values.

Ionization chambers have been designed to obtain the quality fac-

tor in mixed radiation fields based on columnar recombination of

high-LET radiation (Zielczynski, 1962). Measurements are made

over a range of voltages from 2 to 1,200 V. The data are used to

determine the change (slope) of the measured current as a function

of voltage. This slope is related to the quality factor through an

empirical formula (Baarli and Sullivan, 1965). Recombination

94 / APPENDIX A

chambers have been used to obtain the quality factor for mixed

fields at high-energy particle accelerators (Cossairt and Elwyn, 1987;

Hofert and Raffnsoe, 1980). It is not known if the empirical relation-

ships will remain the same for space radiations such as a trapped

protons or GCR. It is also imperative that the dose rate remains

constant while sweeping the voltage, which can take up to an hour

at low dose rates. This may be impractical in LEO and therefore a

second ion chamber with constant voltage would be necessary for

normalization.

Appendix B

Computational Methods

The determination of the types and energy distributions of parti-

cles, transmitted through a shield material and through astronaut

tissues, requires the solution to radiation transport equations of the

process with appropriate boundary conditions for the external space

radiation environment. The relevant transport equations are the

linear Boltzmann equations derived on the basis of conservation

principles (Wilson et al., 1991) for the fluence &i(x,�,E) of particles

of type i moving in direction � with energy E as:

� � �&i(x,�,E) � ��ik(�,��,E,E�) &k(x,��,E�) d�� dE��

� �i(E) &i(x,�,E), (B.1)

where �i(E) is the rate of removal of particles of type i with energy

E moving in direction �, �ik(�,��,E,E�) are the media macroscopic

cross sections for various atomic and nuclear processes k producing

particles of type i with energy E moving in direction �, including

spontaneous disintegration. In general, there are hundreds of fluence

terms [&i (x,�,E)] with several thousand cross-coupling terms

[�ik(�,��,E,E�)] through the integral operator in Equation B.1. The

total cross section [�i(E)] with the medium for each particle type of

energy E may be expanded as:

�i(E) � �i,at(E) � �i,el(E) � �i,r(E), (B.2)

where the first term refers to collision with atomic electrons (at),

the second term is for elastic nuclear scattering (el), and the third

term describes nuclear reactive processes (r). The microscopic cross

sections and average energy transfer are ordered as follows:

�i,at(E) � 10�16 cm2 for which �Eat � 102 eV; (B.3)

�i,el(E) � 10�19 cm2 for which �Eel � 106 eV; and (B.4)

�i,r(E) � 10�24 cm2 for which �Er � 108 eV. (B.5)

95

96 / APPENDIX B

This ordering allows flexibility in expanding solutions to the

Boltzmann equation as a sequence of physical perturbative approxi-

mations. It is clear that many atomic collisions (�106) occur in a

centimeter of ordinary matter but transferring little energy per colli-

sion, whereas �103 nuclear coulomb elastic collisions occur per centi-

meter but with significant energy transfer. In distinction, nuclear

reactions are separated by a fraction to many centimeters depending

on energy and particle type but with large energy transfers. Atomic

interactions are treated with an energy moments expansion in which

the first term is related to LET and higher order moments are related

to range straggling. The secondary electrons form the energy deposit

around the ion path. The elastic scattering is treated as lateral

diffusion. Special problems arise in the perturbation approach for

neutrons for which �i,at(E) � 0, and the nuclear elastic process

appears as the first-order perturbation. The neutron-coupling prob-

lem has been the recent focus of research (Clowdsley et al., 2000)

and further described below along with the reactive processes.

The double differential particle production and fragmentation

cross sections [�ik(�,��,E,E�)] of Equation B.1 are separated into

an isotropic contribution and a remainder as:

� � �F � �iso , (B.6)

where the remainder �F consists of only forward directed secondary

particles and �iso is dominated by lower energy particles produced

in the reaction. The low-energy charged particles are of short range

due to atomic interaction processes and may be solved analytically

(Wilson et al., 1991) but the low-energy neutrons have no atomic

interactions and require a different solution technique (Clowdsley

et al., 2000). The solution to Equation B.1 can likewise be separated

into two parts for which �F appears only in Equation B.1 with solution

&F and a second equation in which �iso appears in Equation B.1 but

with source terms from coupling to the &F field through �iso. The

solution to Equation B.1 for &F can be written in operational form as:

&F � GF &B , (B.7)

where &B is the inbound field at the boundary, and GF is the Green’s

function associated with �F, which reduces to a unit operator on

the boundary denoted by '. Standard practice in space radiation

protection replaces ' as required at some point on the boundary and

along a given ray by the corresponding 'N evaluated for normal

incidence (N) on a semi-infinite slab (Wilson et al., 1994). The errors

in this approximation are made only for the &F field and are second

order in the ratio of beam divergence and radius of curvature of the

object, rarely exceed a few percent for space radiations, and are

COMPUTATIONAL METHODS / 97

always conservative (Wilson et al., 1994). The replacement of ' by

'N as a highly accurate approximation for space applications has the

added advantages that 'N is the natural quantity for comparison

with laboratory experiments and has the following property: if 'N is

known at a plane a distance xo from the boundary (assumed at the

origin), then the value of 'N at any plane x � xo is:

GN(x) � GN(x � xo) GN(xo). (B.8)

Setting x � xo� h, where h is small and of fixed step size gives rise

to highly efficient marching procedure of the radiation transport

code HZETRN (Wilson et al., 1997). Allowing x and xo to take arbi-

trary values gives rise to the nonpertabative solution methods. There

remains the evaluation of the remainder terms �iso of Equation B.1,

especially the low-energy neutron transport.

The remainder of Equation B.1 following the separation given by

Equation B.6 is:

� � �&i(x,�,E) � ��iso,ik(E,E�) &k(x,��,E�) d�� dE�)�

� �i(E)&i(x,�,E) � gi(E,x), (B.9)

where the source term [gi(E,x)] results from the collisional �iso source

with the &F field. As mentioned above, the charged particle fields

of Equation B.9 can be solved analytically leaving the low-energy

neutrons fields to be evaluated. Energy multigroup methods are

currently used for evaluation of Equation B.9 for the neutron fields

(Clowdsley et al., 2000). The solution for the diffuse component

depends on the surrounding material and the accumulation of the

neutrons produced in and scattered from the massive material of

human occupied spacecraft is an essential part of solution of the

Boltzmann equation. The diffuse charged particle component is sim-

pler since the atomic interactions tend to dominate in the appropriate

energy range.

The extent of the nuclear interaction cross section database

required for the transport of cosmic rays spans most nuclear-reaction

physics from thermal energies to energies above tens of GeV n�1,

including a large number of projectile and target material combina-

tions. The types of cross sections required for the transport involve

total yields and secondary energy spectra for one-dimensional trans-

port and double differential cross sections in angle and energy for

three-dimensional transport. Fortunately, neutron and proton cross

sections have been studied at some length in the past. Nuclear-

reaction modeling is required, especially for light and heavy ion

projectiles, to understand the basic physical processes, and to extra-

polate the limited, available experimental data between projectile

98 / APPENDIX B

energies and projectile-target combinations. The usual approach to

database generation is the use of Monte Carlo models or hydrody-

namic models with limited usefulness and success. The best approach

has been to develop solution procedures of the basic quantum

mechanics (Cucinotta et al., 1993; Wilson, 1974).

A microscopic theory for the description of nuclear fragmentation is

being developed through the study of the summation of the nucleus-

nucleus, multiple-scattering series for inclusive reactions where a

single reaction species is considered. This approach originated in a

theory for high-energy alpha-particle fragmentation (Cucinotta

et al., 1993) and has been extended to recast the abrasion-ablation

model in microscopic form (Cucinotta and Dubey, 1993). The micro-

scopic theory represents a unified approach where a single formalism

generates all the cross sections required for heavy-ion transport.

The evaluation of the microscopic theory requires the interaction

amplitudes among basic nuclear constituents. The strong interaction

and quantum phase space limitations (Pauli principle) modify the

amplitudes for bound constituents and have important impact on

elastic and reactive processes (Tripathi et al., 1999). Basic nuclear

theory and media modified interaction amplitudes with the optical

theorem can now, for the first time provide high quality total and

reactive cross sections as recently incorporated into the Langley 90

database (Tripathi et al., 1998; 1999) and have been adopted by other

groups and agencies and stand ready for application to the HZE

fragmentation process. A recent blind test in iron fragmentation

experiments of various models has been very encouraging for future

database generation (Cucinotta et al., 1998; Zeitlin et al., 1997).

Appendix C

ThermoluminescenceDosimetry Materials

The more usual TLD materials for passive personal radiation

dosimetry have been reviewed and described by McKeever et al.

(1995). For dosimetry in LEO, two materials have been used most

often, namely, lithium fluoride, doped with magnesium and titanium

(LiF:Mg,Ti) and calcium fluoride, doped with thulium (CaF2:Tm).

C.1 Lithium Fluoride, Doped with Magnesium

and Titanium

Radiation dosimetry-quality lithium fluoride generally comes in

three forms. The differences between them lie in the isotopic abun-

dance of lithium, thus: natural abundance (92.5 percent 7Li and 7.5

percent 6Li), enriched in 6Li (95.5 percent), and enriched in 7Li (99.5

percent). Such detectors are available from a variety of manufactur-

ers. Perhaps the most widely known forms are known as TLD-100�,

TLD-600� and TLD-700�, respectively, using the original trade

names of the old Harshaw Chemical Company (now Saint-Gobain

Industries). Each of these is doped with magnesium, titanium and

hydroxyl impurities. Since 6Li has a large cross section for neutron

interaction, via the reaction 6Li(n,�)3H, then 6Li-enriched versions

(e.g., TLD-600�) are more sensitive to thermal neutrons (below

�1 keV) than either of the other versions (such as TLD-100� or

TLD-700�).

The usual procedure for using LiF TLDs is to anneal the sample at

400 °C for 1 h. This is then followed by either a slow cool to 80 °C, and

a hold at this temperature for 24 h, or a slow cool to 100 °C, and a

hold at this temperature for 2 h (although other similar recipes are

often used) (Furetta and Weng, 1998; Horowitz, 1984b; McKeever,

99

100 / APPENDIX C

1985; McKeever et al., 1995; Niewiadomski, 1993). Reproducible con-

trol of the temperature and the timing of the heating and cooling cycles

are considered to be critical. The purpose of the heating schemes is

two-fold: (1) to reset the defect equilibrium state within the samples

after each use, and (2) to empty charge from stable defects which may

remain after the first irradiation and readout cycle. To ensure stable

resetting of the defect equilibrium the lengthier, but more reliable,

400 °C for 1 h plus 80 °C for 24 h is often recommended (McKeever

et al., 1995).

A typical glow curve for LiF:Mg,Ti (in this case TLD-100�) treated

by annealing at 400 °C for 1 h plus 100 °C for 2 h is shown in

Figure C.1. Although these data were obtained with the faster 400 °C

for 1 h plus 100 °C for 2 h anneal, they serve to demonstrate the

complexity of the glow curve. The sample has been irradiated with

1.5 MeV protons (1.8 � 1010 protons cm�2). The glow curve is a

complex overlap of several individual signals. The figure shows two

methods for deconvoluting the signal into its component parts. No

single deconvolution model has yet been agreed upon for the glow

curve from LiF:Mg,Ti but, nevertheless, the data of Figure C.1 show

the complexity of the glow curve with its multiple components. Glow

curves for other versions of LiF:Mg,Ti (e.g., TLD-600� and TLD-

700�) are similar.

The TL signal corresponding to Peak 5 (Figure C.1) is usually

taken to be the signal of interest. The response of LiF:Mg,Ti to

absorbed dose from gamma rays indicates a linear-supralinear

behavior, with saturation occurring near 103 Gy (McKeever et al.,

1995). Supralinearity of the signal starts at a few gray (McKeever

and Horowitz, 1990). The dose response of this material to all low-

LET (i.e., �10 keV �m�1) radiation is essentially identical to that

of 60Co or 137Cs irradiation. Likewise, the dose responses of other

versions of LiF:Mg,Ti (e.g., TLD-600� and TLD-700�) are generally

considered to be similar to that of TLD-100�.

The response of the material to heavier charged particles, however,

changes considerably from that resulting from irradiation with low-

LET irradiation. The increased ionization density around the track

of the incident particle results in a decrease in sensitivity and a

decrease in supralinearity for Peak 5, but with a concomitant

increase in sensitivity and supralinearity for the higher temperature

peaks (i.e., Peaks 6 to 10 in Figure C.1). The change in the shape of

the LiF glow curve as a function of LET of the particle is represented

in the data of Figures C.2 and C.3.

The normal approach for determining the absorbed dose from a

radiation source (photon or particle) using LiF:Mg,Ti dosimeters is

C.1 LiF:Mg,Ti / 101

Fig. C.1. TLD-100� glow curve for 1.5 MeV protons (1.8 � 1010 protonscm�2) following an anneal treatment of 400 °C for 1 h plus 100 °C for 2 h.The heating rate was 1 °C s�1 (Aviles et al., 1999). The figure shows theresults of two different methods [(a) and (b)] for deconvoluting the signalinto its component parts.

to calibrate the Peak 5 signal against absorbed dose from gamma

rays (e.g., using 137Cs or 60Co).11 Thus, one determines the gamma-raydose from 137Cs or 60Co that is required to produce the same signalintensity as the incident radiation field (i.e., one determines the

11 For first-order TL processes (as in LiF) the peak area is proportional to peakheight. The best accuracy these days is achieved with glow curve deconvolution inorder to separate the individual responses of the individual peaks (Horowitz andYossian, 1995). This is particularly important since there are multiple overlappingpeaks in TLD materials, each of which has a slightly different response to the variouscharged particles expected in a space environment.

102 / APPENDIX C

Fig. C.2. Change in the shape of the glow curve (from TLD-600�) as a

function of charge particle type. ‘‘HTR’’ is the high temperature ratio and

is the integrated area under the curve, between the limits shown, divided

by the height of Peak 5. The data in this Figure have been normalized to

the maximum of Peak 5 to facilitate calculation of HTR. The irradiations

(from top to bottom) are: 241Am alpha, 174 keV �m�1; thermal neutrons,

130 keV �m�1; 19F ions, 102 keV �m�1; 19F ions, 30 keV �m�1; 12C ions,

12.8 keV �m�1; and 60Co gamma rays (Schoner et al., 1999).

equivalent 137Cs or 60Co gamma-ray dose in gray). It was noted above

that the sensitivity of Peak 5 decreases with increasing LET. For a

single species of radiation, one thus needs to account for this depen-

dence in order to determine the equivalent gamma-ray dose. In Fig-

ure C.4 we see the change in the efficiency of Peak 5 (for TLD-700�)

as a function of LET. In this case, the normalization dose is obtained

from 60Co. To be noted is an over-response caused by high-energy

protons with respect to the gamma-ray dose, and a monotonic decrease

in sensitivity for LET greater than approximately 10 keV �m�1.

C.1 LiF:Mg,Ti / 103

Fig. C.3. High temperature ratio (HTR) plotted against LET (L�) for

TLD-100� (●), TLD-600� (�), and TLD-700� (�) (Schoner et al., 1999).

Similar data have been published elsewhere (e.g., Horowitz, 1984b;

McKeever, 1985; Yasuda and Fujitaka, 2000). The proton over-

response indicated in Figure C.4 is �10 percent, which is about

the same as that reported elsewhere (Horowitz, 1984b; Yasuda and

Fujitaka, 2000). However, larger over-responses have been reported

[e.g., Schoner et al. (1999) report a proton over-response of �20 percent].Benton et al. (2000) define the efficiency function �(L) in three

regions of LET, thus:

�(L) � 1, (C.1a)

for L � 2 keV �m�1;

�(L) � a1 � b1(log L) � c1(log L)2, (C.1b)

for 2 � L � 50 keV �m�1; and

�(L) � a2 � b2(log L) � c2(log L)2, (C.1c)

for L � 50 keV �m�1;

104 / APPENDIX C

[ ]

L

L

Fig. C.4. Response of Peak 5 from TLD-600� and TLD-700� to dose-

averaged LET (Equation C.1). The response is normalized to 60Co gamma

rays, and is therefore defined as the TL efficiency [�(L)] with respect to 60Co

(adapted from Benton et al., 2000).

where a1 � 0.9650, b1 � 0.1805, c1 � 0.2100, a2 � 1.7536, b2 �

0.8060, and c2 � 0.1016. This ignores the proton overresponse at low-

LET. From the available published data the conclusion is reached

that this material does indeed overrespond to high-energy protons.

However, some of the data are equivocal, and the extent of the over-

response is not clear, differing by several percent from report to report.

Important issues include the precise batch of LiF used, the exact

annealing conditions used (temperatures, duration, cooling rates, etc.)

and the region of interest used in the determination of the TLD

response (i.e., integration between temperature X and Y, peak height

calculation, etc.) and, of course, the reliability of the dosimetry in the

individual reports.

Since the detailed TLD response is a function of the pattern of

energy deposition around the particle track (i.e., the track structure),

C.2 CaF2:Tm / 105

the precise behavior will be dependent upon both particle energy (E)

and particle charge (Z). In particular, the radius of the track

increases as Z increases and thus two particles having the same E

but different Z are not expected to produce the same value of �(L).

Data by Avila et al. (1999) and Geiss et al. (1998) on the efficiency

of TL from LiF:Mg,Ti for several particles as a function of both E

and Z illustrate the difficulty of providing a simple correction for

the E,Z-dependence. An effective means to deal with this through

calibration of TLDs for each particle type is not trivial since the

number of different types of charged particle in the space radiation

field is too large. However, an algorithm to calculate the efficiency

(�) as functions of both E and Z is described by Geiss et al. (1998)

and this may prove to be a promising approach.

For space flight, particles of Z � 26 are of less importance (Benton

et al., 2000). Also to be noted is a reported ‘‘non-universality’’ of

�(L) for a given material (Horowitz, 1984b). Variations in the defect

structure (impurity content, defect equilibrium, etc.) lead to varia-

tions in the exact shape of the �(L) curve. Reproducibility in the

heating and cooling cycles during annealing is considered essential

and calibration of each TLD batch may be necessary in critical

situations.

C.2 Calcium Fluoride, Doped with Thulium

Calcium fluoride comes in several forms for use in TL dosimetry,

but the form of interest for space dosimetry applications is CaF2:Tm,

known as TLD-300�. The material is generally considered to be

reusable without any thermal annealing treatments if used for

absorbed doses �200 mGy, although annealing at �400 °C for at

least 30 min is recommended for low-dose (�30 mGy) applications

(McKeever et al., 1995).

The TLD-300� glow curve is shown in Figure C.5. As with LiF,

the glow curve consists of multiple overlapping peaks. The main

signals used in dosimetry are Peak 3, and Peaks (5 � 6) together

(Figure C.5b). The gamma-ray dose response shows linear behavior

for Peak 3 up to �10 Gy, followed by saturation. Peak 5, however,

shows linearity up to �1 Gy, followed by supralinearity, with maxi-

mum supralinearity at �105 Gy. This difference in behavior in the

responses of Peaks 3 and 5 to gamma-ray dose results in a markedly

different behavior for charged-particle irradiation. Specifically, the

ratio of Peaks (5 � 6) to Peak 3 increases with increasing LET, as

Figure C.6 illustrates for the TLD-300� glow curve after irradiation

106 / APPENDIX C

Fig. C.5. Two glow curves from TLD-300�, analyzed using either

(a) seven peaks, or (b) six peaks. The six-peak structure is usually adopted

in dosimetry (Bos and Dielhof, 1991).

C.3 LiF:Mg,Cu,P / 107

Fig. C.6. TLD-300� glow curve following irradiation with 241Am alpha

particles (Buenfil et al., 1999).

with 241Am alpha particles. The variation in the heights of the glow

peaks with increasing LET is shown in Figure C.7.

It is clear that the peak-height ratio [Peaks (5 � 6)/Peak 3] is a

sensitive indicator of LET of the charged-particle field to which the

sample is exposed. The efficiency with respect to gamma-ray irradia-

tion �(L) decreases with increasing LET in a similar fashion to that

for LiF. The effect of different Z on the efficiency curve has not been

studied, nor has the universality of the response yet been thoroughly

investigated.

C.3 Lithium Fluoride, Doped with Magnesium, Copper

and Phosphorus

The glow curve for LiF:Mg,Cu,P is shown in Figure C.8. The mainpeak shows a linear response to gamma-ray irradiation, up to�10 Gy, without any supralinearity. For high-LET charged-particleirradiation the glow curve shows very low sensitivity (approximatelya factor of 100 lower than LiF:Mg,Ti) with a pronounced highertemperature peak (Peak 5) compared with the main dosimetry peak

108 / APPENDIX C

Fig. C.7. Variation in the heights of Peaks 3 and 5, normalized to the

heights of these peaks after 60Co gamma-ray irradiation, for various heavy

charged particles (HCP) (RHCP/RCo-60). The arrows indicate the entrance and

exit LET values for the particles indicated. (�) helium ions, () neon ions,

(�) 0.7 MeV protons, () 5.3 MeV alpha particles, (●) 120 MeV 12C ions. The

maximum fluence was 1011 particles cm�2 (Buenfil et al., 1999).

(Peak 4 in Figure C.8.) The low sensitivity is a result of the microdosi-metric distribution of trapped charge along the track of the chargedparticle and the rapid saturation of the TL signal at high ionizationdensities (Horowitz and Stern, 1990). The sensitivity to gamma raysand low-LET irradiations, however, is considerably higher than thatof LiF:Mg,Ti. Thus, this material acts as an efficient discriminatorin mixed low- and high-LET fields. However, a detailed study of the�(L) function for this material has not yet been performed.

C.4 Aluminum Oxide, Doped with Carbon

Figure C.9 shows the glow curve for Al2O3:C irradiated by gammarays. The main peak shows a linear response to gamma-ray irradi-ation, up to �50 Gy, without any significant supralinearity. The

C.4 Al2O3:C / 109

TL (

arb

itra

ry u

nits) 1

0.5

Temperature ( K)

300 400 500 600

6

532

01

º

Fig. C.8. Glow curve from gamma-ray irradiated LiF:Mg,Cu,P (McK-

eever et al., 1993).

TL

(a

rbitra

ry u

nits)

0.5

0.9

0.8

0.7

0.6

0.4

0.3

0.2

0.1

0100 150 200 250 30050

1

Temperature ( C) º

Fig. C.9. Glow curve from gamma-ray irradiated Al2O3:C.

efficiency of Al2O3:C for the detection of gamma rays is claimed tobe as high as 40 to 50 with respect to LiF:Mg,Ti (McKeever et al.,1995). Under alpha-particle irradiation, the main peak saturatesat about the same level, but a smaller, higher temperature peaknear 325 °C grows linearly, then supralinearly, before saturation(Moscovitch et al., 1993). The TL efficiency � to alpha particles, withrespect to gamma rays, is more than an order of magnitude less forAl2O3:C compared to LiF:Mg,Ti (Mukherjee and Lucas, 1993). This

110 / APPENDIX C

is consistent with the differences observed in the gamma-ray dose

response and suggests that this material (like LiF:Mg,Cu,P) would

be very effective at discriminating against high-LET particles in

mixed low-LET/high-LET fields.

A disadvantage of Al2O3:C for TL dosimetry is an observed heating

rate dependence due to thermal quenching of the luminescence effi-

ciency (Akselrod et al., 1998). This disadvantage can be overcome,

however, by using this material as an OSLD instead of a TLD. Here

one measures the radiation-induced luminescence signal using light

from a laser beam to stimulate the luminescence. Several OSL read-

out modes have been suggested and successfully developed (Akselrod

and McKeever, 1999; Botter-Jensen et al., 1997).

A complete description of �(L), using either TL or optically stimu-

lated luminescence, for this material is not yet available.

C.5 Evaluation of Absorbed Dose and Dose Equivalent

C.5.1 Equivalent Gamma-Ray Dose

For a linear growth of a TL signal with gamma-ray dose (from60Co or 137Cs):

TL � D, (C.2)

where is a proportionality constant describing the efficiency of TL

production for gamma-ray irradiation (in TL units per unit gamma-

ray dose). For linear growth of TL with absorbed dose from a particle

with linear energy transfer L:

TLL � � D, (C.3)

with � � 1.

For a mixed particle field of fluence &(L), the delivered absorbed

dose to TLD is �D(L)dL, summed over all angles and all times. Here,

D is the absorbed dose delivered to TLD by particles of linear energy

transfer L and is a function of L. Thus:

TLL � � �(L) D(L) dL. (C.4)

The desired quantity at this stage is D � �D(L) dL. The difficulty

is that TLL is a single number whereas �(L) and D(L) are functions.

Thus, the problem is ‘‘ill-posed.’’ The simplest, and least satisfactory,

solution is to determine that gamma-ray dose which gives rise to

the same TL signal as that due to the particle irradiation. That is,

C.5 EVALUATION OF ABSORBED DOSE / 111

one uses D for the absorbed dose in TLD due to the charged-particle

field, where:

D � � �(L) D(L) dL. (C.5)

C.5.2 Mean, or Effective, Linear Energy Transfer

An improved approach is to independently measure the LET flu-

ence spectrum [&(L)] and obtain an effective efficiency (�) determined

from either a mean or an effective LET (Li) suitably averaged over

some i. For example, from the known &(L) one can obtain a dose-

averaged LET, thus:

LD �� D(L) L dL

� D(L) dL, (C.6)

or a fluence-averaged LET, thus:

L& �� &(L) L dL

� &(L) dL. (C.7)

One can use the value of Li to determine � from an experimentally

measured function �(L) for the specific TLD being used. Alterna-

tively, one can define a mean efficiency from:

� �� (L) D(L) dL

� D(L) dL, (C.8)

using an experimentally measured �(L) curve (e.g., Figure C.4).

Using �, one then approximates Equation C.5, thus:

D � � �(L) D(L) dL � � � D(L) dL, (C.9)

and obtains the equivalent gamma-ray dose delivered by the particle

field from:

D �D

�. (C.10)

An effective LET for TL can also be evaluated using either the

LiF or CaF2 glow curves. Since the TL glow curve from a mixed

particle field can be viewed as the superposition of TL glow curves

due to individual particles [i.e., TL � � TL(L) dL] then one may use,

for example, the LiF (TLD-100�) glow curve to determine the HTR

value for the mixed particle space radiation field and compare this

with Figure C.3 to arrive at an ‘‘effective’’ LET (LTLD) (Schoner et al.,

112 / APPENDIX C

1999; Vana et al., 1996). The glow curve from CaF2 (TLD-300�)

(Figure C.7) can be used in the same way. A question arises, however,

concerning how well the effective LET calculated in this way com-

pares with the mean LET calculated using Equations C.6 or C.7.

Yasuda and Fujitaka (2000) demonstrate considerable differences

between LTLD and LD (4.6 and 14 keV �m�1, respectively) using a

typical space radiation field spectrum as estimated from data from

a TEPC (Badhwar et al., 1996a). Nevertheless, using an effective

LET value determined in this way, one can also determine an effec-

tive efficiency �TLD from the known �(L) function and estimate the

equivalent gamma-ray dose using Equation C.9.

C.5.3 Dose Equivalent

The quantity of interest is H (in sievert) determined from:

H � Q D. (C.11)

Q is a function of LET and for a spectrum of charged particles in a

mixed particle field:

H � � Q(L) D(L) dL. (C.12)

Using the calculated equivalent gamma-ray dose in TLD (Equation

C.9), H can be approximated by:

Hi � Qi � D(L) dL � Qi D � Qi

D

�, (C.13)

where the subscript i refers to either D (dose-averaged) or & (fluence-

averaged), and Qi can be determined from ICRP (1991) using the

mean or effective LET value (Li) determined from one of the above

methods.12

12 From an analysis based on photon interaction coefficients, tabulations of electron

and proton stopping powers, and data on muon stopping powers, it is concluded that

for a 137Cs calibration in terms of tissue kerma, the TLDs (LiF) will give an estimate

of absorbed dose to a small mass of tissue for all low-LET components to within five

percent. The energy ranges taken into account include the majority of the low-LET

contributions to dose—for photons up to 10 MeV there would be less than a five

percent overestimate; for electrons up to 50 MeV, an estimate within two percent;

for protons from a few MeV up to 2 GeV an estimate within two percent if only coulomb

interactions are considered; for the cosmic radiation secondary muon spectrum, there

would be an underestimate of about five percent. In these calculations, it was assumed

that the relative light conversion efficiencies for energy deposition by these radiations

in the 7LiF:Mg,Ti remained close to unity (Bartlett et al., 2000).

C.5 EVALUATION OF ABSORBED DOSE / 113

Using the TEPC fluence data of Badhwar et al. (1996a) one gets

QTLD � 1.3 if one uses LTLD from the LiF HTR method, or QD � 2.5

if one uses LD calculated directly from the TEPC data (Yasuda and

Fujitaka, 2000). However, difficulties with such comparisons arise

if there is a significant neutron component and if LiF has a low

response to these components, and/or there is a different secondary

charged-particle spectrum compared to that in tissue.

Yasuda (2001) combined LET dependencies of the low temperature

and high temperature TL peaks from TLD-700� to provide a function

�HL(L) to mimic the Q(L) function and thereby calculate H from:

H � (1 � ) XHT � XLT, (C.14)

when

�HL(L) � ( � 1) �HT (L) � �LT (L), (C.15)

where �HL (L) and �LT(L) are the TL efficiencies of the high tempera-

ture (HT) and low temperature (LT) peaks, respectively; XHT and XLT

are the TL values for the high temperature and low temperature

TL glow peaks, respectively; and is an empirically determined

constant selected to satisfy the relationship �HL(L) � Q(L) over the

entire range of L. There are difficulties associated with this approach

if there is a significant neutron component of the radiation field.

An alternate approach is to limit the use of TLDs to the measure-

ment of absorbed dose from the low-LET (�10 keV �m�1) region for

which Q � 1. In this region, H � QD � D. To do so one needs to

employ a TLD material with excellent discrimination between high-

and low-LET particles. As already described, LiF:Mg,Cu,P or Al2O3:C

may be suitable, but LiF:Mg,Ti or CaF2:Tm would be unsuitable.

One should still take account of differences between absorbed dose

to TLD and absorbed dose to tissue.

To determine H from the high-LET region of the incident field

spectrum one could use a PNTD. Thus, if information is available

about the LET spectrum &(L), then one can calculate H from

H � � D(L) Q(L) dL (Appendix D) and the total H would then be

given by a combination of TLDs and PNTDs, thus:

H � (DQ)TLD � �� D(L) Q(L) dL�PNTD

(C.16)

To provide the best discrimination between the low- and high-LET

regions of the incident spectrum, PADC/CR-39� would be a suitable

114 / APPENDIX C

PNTD material. Since there would be some overlap between the

two detectors each device (TLD and PNTD) would have to be fully

characterized as a function of LET and corrections made to avoid

‘‘double counting’’ the absorbed dose in the overlapping region.

Appendix D

Plastic Nuclear TrackDetectors

D.1 Description of Method

PNTDs register charged particles by means of etchable damage

to the detector structure. The damage trail in a material, which

constitutes a charged-particle track, is a result of local deposition of

energy during the passage of the particle and as such may be related

to restricted LET (L�) or to lineal event density (Paretzke, 1977) or

to radial dose (Butts and Katz, 1967) (see also discussion of Paretzke,

1982). The damage is generally permanent but may be partly

repaired or may be modified over time, influenced by factors such

as temperature, humidity, and the local presence of oxygen or other

gases. The particle tracks, the damage ‘‘trails,’’ may be viewed

directly with an electron microscope in some instances (Price and

Walker, 1962; Silk and Barnes, 1959) or may be rendered visible

under an optical microscope by etching with a suitable solvent

(Young, 1958). The track is developed to form a pit of approximately

conical shape by the preferential dissolution along the particle track,

with the condition that the track etch rate vT is greater than the

bulk etch rate vB. The resulting conical etched pits have entrance

diameters in the range of a few tenths to a few tens of microns. For

an observable track to be formed vT /vB must exceed some threshold

value and this in turn is related to L� or similar parameter of the

particle. The value of vT /vB will also determine the maximum ‘‘accep-

tance’’ angle of incidence for tracks to be made visible. For different

particle parameters, the different values of vT will result in different

evolutions of the etched pit shape with time of etching for a given

set of etchant parameters (chemical composition, molarity and tem-

perature). The measurement of the dimensions of etched pits and

their dependence on etching time allows an estimation to be made

115

116 / APPENDIX D

of the charged particle’s L� and residual range, and, not always

unambiguously, allows identification of particle charge and energy.

Stacks of thin sheets of detector material are used to trace the paths

of particles. An introduction to the more general study of track etch-

ing and its many applications may be found, for example, in Durrani

and Bull (1987) and Fleischer et al. (1975).

D.2 Neutron Dosimetry

Until the late 1970s, neutron detection using PNTDs was only

possible by means of (n,�) or (n,f) reactions and for neutron energies

above about 1.5 MeV by the detection of recoil nuclei. The situation

changed with the discovery of the proton response of the commonly

used plastic, PADC/CR-39�13 (Cartwright et al., 1978; Cassou and

Benton, 1978). Other track etch materials, notably some forms of

cellulose nitrate are also able to register protons, but with low sensi-

tivity. Depending on the etch process used, protons of energies up

to about 10 MeV may be detected by PADC/CR-39�, corresponding

to LETs (in water) down to about 5 keV �m�1.

The application of PADC/CR-39� to neutron personal dosimetry

was recognized and demonstrated by several authors (Al-Najjar

et al., 1979; Benton et al., 1980; Griffith et al., 1981; Ruddy et al.,

1981; Somogyi and Hunyadi, 1980). The addition of a converter layer

for the neutron energy region from thermal to a few kiloelectron

volts (Bartlett et al., 1986) produces a dosimeter with acceptable

energy dependence of response for most practical purposes. The early

widespread application however, was hampered by problems of plas-

tic consistency. There have been recent improvements in the methods

of material manufacture (Ahmad and Stejny, 1991).

For neutron dosimetry applications, the counting of low track den-

sities can be accomplished for etched tracks in thin films by rapid

readout techniques using highly colored detector films or electroni-

cally by the development of spark counting. The process of electro-

chemical etching (Sohrabi, 1974; Tommasino, 1970; Tommasino and

Armellini, 1973) greatly increases the size of the etched pits such

that they are visible to the unaided eye and can be counted automati-

cally by simple low-power optical systems. In electrochemical etching,

13 Polyallyl diglycol carbonate is the polymeric form of allyl diglycol carbonate, which

is the commercial name for the chemical diethylene glycol bis (allyl carbonate), also

known as 2,2�oxydiethylene bis (allyl carbonate) or as the oxydi-2, 1-ethanedyl d-2-

propanyl diester of carbonic acid.

D.3 DOSIMETRY OF COSMIC RADIATION FIELDS / 117

a high alternating electric field is applied across the detector foil

during chemical etching. The field strengths are typically in the

range of 20 to 50 kVRMS cm�1 at frequencies in the range 50 Hz to

10 kHz. For pits, which attain a sharply pointed shape as the chemi-

cal etching proceeds, the field strength is enhanced at the tip and

local breakdown of the material leads to catastrophic damage. The

characteristic ‘‘tree’’ shaped formation produced has a diameter in

the range of 10 to several hundred microns compared to chemically-

etched pit diameters of a few tenths to a few tens of microns. The

electrochemically-etched pits are visible to the naked eye and can

be counted by simple automated low magnification optical systems.

A variation on this process is electro-etching, or ‘‘blow up’’; in which

small pits which have been produced by chemical etching, or rela-

tively underdeveloped pits produced by low-frequency electric field

electrochemical etching, or a combination of both, are subjected to

a short accelerated high-frequency electro-etch. Electrochemical

etching is sometimes preceded by a short chemical etch.

Detailed descriptions of fundamental aspects of neutron dosimetry

using PNTDs, including information on the different etch processes

and dosimetric characteristics may be found in two review papers

(Harrison and Tommasino, 1985; Tommasino and Harrison, 1985),

and in a recent survey of the status of development in Europe (Harvey

et al., 1998).

D.3 Dosimetry of Cosmic Radiation Fields

PNTDs, in particular, PADC/CR-39�, can be used to determine the

neutron and HZE components, and, in part, the proton component of

the cosmic radiation field in spacecraft. The material type, and in

some cases the batch, needs to be calibrated (see calibration curves,

adapted from Keegan, 1996, Figure D.1) either for a sample from

the batch used, or by a combination of an a priori calibration curve

for a material plus an in-flight self-calibration (Gunther et al., 1999;

Wiegel et al., 1988). The in-flight self-calibration takes account of

changes in sensitivity of the detector owing to flight conditions. Full

etch track analysis allows the determination of HZE particle fluence

and fluence rate, but is extremely time consuming unless automated.

The LET threshold of PADC/CR-39� is in the range 5 to 10 keV �m�1,

enabling protons of energy up to about 10 MeV to be detected directly,

as well as HZE particles. Secondary charged particles from nuclear

(strong force) interactions of higher energy protons and of neutrons,

in the detector material itself or surrounding material, are also

118 / APPENDIX D

vT/ v

B

10 100 10001

2

1

3

4

5

6

7

8

9

10

L200 in PADC/CR-39® (keV µm–1)

Type 2 data

Type 1 data

(C, Ca and Fe ions)

(C ions)

Fig. D.1. Calibration curves for two types of PADC/CR-39� (adapted

from Keegan, 1996). L200 is the restricted LET (L�; � � 200 eV) and

vT /vB is the track etch rate divided by the bulk etch rate.

detected. The full analysis of the particle track parameters allows the

determination of LET spectra (e.g., Figure D.2) (Badhwar et al., 1996a;

1996b; Benton and Parnell, 1988; Doke et al., 1995; Wiegel et al.,

1988; Yasuda et al., 2000). The dimensions of a large number of tracks

of particles, mostly secondary protons, are measured and their LET

determined via the calibration curve [samples of PADC/CR-39� are

characterized in terms of LET (in water) for proton and heavy charged-

particle beams for the LET range from about 5 to 1,000 keV �m�1].

By a measurement of the distribution of absorbed dose (fluence times

LET) in LET, H can be determined. The value of H so determined is

approximate; the spectrum of secondary charged particles generated

in the detector and its immediate surroundings will be different from

that for tissue. The known dependence of response of the detector on

angle of incidence may be corrected for by assuming that the radiation

field is isotropic, or rendered so by the movement of the detector if

worn. This assumption must be made with care (for example, Benton

and Parnell, 1988). This technique has been successfully applied, for

example, to the measurement of H in the radiation field in spacecraft

(Doke et al., 1995; Yasuda et al., 2000), in supersonic aircraft at flight

altitudes and the European Organisation for Nuclear Research/Euro-

pean Union reference fields (CERF) simulated cosmic radiation neu-

tron field component (O’Sullivan et al., 1999). An example for CERF,

a description of which may be found in Hofert and Stevenson (1994),

is shown in Figure D.3 (O’Sullivan et al., 1999). An approximation to

H from all field components depositing energy with LET greater than


Inte

gra

l N

um

be

r F

Iux (

cm

–2 s

r –1d

–1)

L∞ in tissue (keV µm-1)

105

104

103

102

101

100

10–1

10–2

10–3

10–4

10–1 100 101 103102

10–4

105

104

102

103

101

100

10–1

10–2

10–3

104

Fig. D.2. LET (L�) spectrum for PADC/CR-39� compared with that

obtained with TEPC (adapted from Badhwar et al., 1996a).

Fig. D.3. Integral dose equivalent as a function of LET (L�) for CERF

top concrete neutron field, using Q(L) from ICRP (1991) (adapted from

O’Sullivan et al., 1999).

120 / APPENDIX D

5 to 10 keV �m�1 is determined by this approach. The contribution

from all field components to H for lower values of LET can be deter-

mined by passive TLDs or OSLDs.

Another, more approximate approach, is to use a series of different

materials each of different LET threshold to obtain particle fluence

in a number of LET ‘‘bins’’ (for example, Kopp et al., 1999). Materials

that can be used are PADC/CR-39�, cellulose nitrate and polycarbo-

nate (Lexan, Makrofol�).

PNTDs may be used to estimate the higher-LET component of the

cosmic radiation field (comprising most of the neutron component,

plus the nuclear interaction component of high-energy protons and

HZE particles) using the simpler, less time consuming techniques

developed for neutron personal dosimetry. Electrochemically etched

pits are identified and counted. Readout procedures are fully auto-

mated. Figure D.4 shows neutron dose equivalent response data for

some operational personal dosimetry systems, three PADC/CR-39�

etched track dosimeters [one chemical etch after Fiechtner and

Wernli (1999) and two electrochemical etches after Bartlett et al.

(2001) and Luszik-Bhadra et al. (1999)], and one electrochemical etch

polycarbonate (Makrofol�) after Jozefowicz et al. (1997). The higher

energy response data were obtained in quasi-monoenergetic neutron

RH

(m

Sv

–1)

1

2

0.5

0.2

0.1

5

E (eV)

105

106

107

108

(d)

(c)(b) (a)

Fig. D.4. Energy dependence of the ambient dose equivalent response

(RH), normalized to unity for americium-beryllium neutrons: for three PADC/

CR-39� etched track dosimeters [one chemical etch (a) adapted from Fiechtner

and Wernli (1999), two electrochemical etches (b) adapted from Bartlett et al.

(2001) and (c) adapted from Luszik-Bhadra et al. (1999)], and one electrochemi-

cal etch polycarbonate (Makrofol�) (d) adapted from Jozefowicz et al. (1997).


fields at the Paul Scherrer Institute for the neutron energy range 20

to 70 MeV and at 160 and 180 MeV at the Svedberg Laboratory,

Uppsala University. The quasi-monoenergetic fields have about 50

percent of the neutron fluence in the peak and the remainder distrib-

uted in approximately uniform lethargy down to thermal (Blomgren,

1997; Schrewe, 1997).14 The lower energy response must be ‘‘stripped

off’’ to obtain the response for the peak neutrons, a procedure that

results in significant uncertainties in the high-energy response data.

The high-energy response data given in Figure D.4 show reasonable

agreement between four different etched track systems. These data

are also broadly in agreement with those obtained for high-energy

protons (Spurny, 1995). HZE particles will also produce an etchable

track and be counted as if produced by a neutron, or neutron-like

iterations of high-energy protons, and the same response factor

applied (about 5 to 10 �Sv per track for a typical electrochemical etch

system), which will overestimate the HZE component of H. However,

an additional chemical etch allows discrimination.

14Blomgren, J. (1997). Personal communication (Svedberg Laboratory, Uppsala Uni-

versity, Uppsala, Sweden);

Schrewe, U.J. (1997). Personal communication (Physikalisch-Techiseche Bunde-

sanstalt, Braunschweig, Germany).

Glossary

absorbed dose (D): The quotient of d� by dm, where d� is the mean energy

imparted to matter of mass dm, i.e., D � d�/dm. The unit for D is joule

per kilogram (J kg�1) with the special name gray (Gy).

administrative level: A predetermined reference value of a quantity, below

a dose limit, that triggers a specified course of action when the value is

exceeded or is expected to be exceeded. In the case of space activities,

NASA sets administrative levels for a given mission.

albedo neutrons: Secondary neutrons produced by interactions of galactic

cosmic radiation and the atmosphere, and reflected back into space.

as low as reasonably achievable (ALARA): A principle of radiation

protection philosophy that requires that exposures to ionizing radiation

should be kept as low as reasonably achievable, economic and social factors

being taken into consideration. The protection from radiation exposure

is ALARA when the expenditure of further resources would be unwar-

ranted by the reduction in exposure that would be achieved. In the case

of space activities, ALARA applies to actions taken to keep all doses to

the astronauts as low as reasonably achievable, balancing the mission

objectives with practical dose reduction steps.

coronal mass ejection: A transient outflow of plasma from or through

the solar corona, which may be associated with the generation of solar

particle events.

deterministic effects: Effects for which the severity varies with dose and

for which a threshold usually exists, e.g., cataracts and skin burns.

dose: A general term used when the context is not specific to a particular

dose quantity. When the context is specific, the name or symbol for the

quantity is used, i.e., absorbed dose (D), mean absorbed dose (DT), dose

equivalent (H), effective dose (E), equivalent dose (HT), or organ dose

equivalent (HT).

dose equivalent (H): The absorbed dose (D) at a point in tissue, modified

by the quality factor (Q) at that point, i.e., H � QD. The unit for dose

equivalent (H ) is the joule per kilogram (J kg�1) with the special name

sievert (Sv).

dose limit: A limit on radiation dose that is applied for exposure to individu-

als or groups of individuals in order to prevent the occurrence of radiation-

induced deterministic effects or to limit the probability of radiation related

stochastic effects to an acceptable level. For astronauts working in LEO,

unique dose limits for deterministic and stochastic effects have been rec-

ommended by NCRP.

dosimetry records: The collections of information that document the radia-

tion doses that astronauts receive from space flight, mission-related

122

GLOSSARY / 123

aviation activities, mission-related biomedical research, and any other

occupational doses. The dosimetry records contain, or are linked to, all

the basic information that is necessary to obtain the radiation doses.

effective dose (E): The sum over specified tissues of the products of the

equivalent dose in a tissue (HT) and the weighting factor for that tissue

or organ (wT), i.e., E � �T

wTHT. Effective dose (E) applies only to stochastic

effects and is expressed in sievert (Sv).

equivalent dose (HT): The product of the mean absorbed dose in an organ

or tissue (DT) (i.e., DT, R for a given type of radiation R), and the radiation

weighting factor (wR) that accounts for the relative biological effectiveness

of the radiation type. For external exposure, wR applies to the radiation

type incident on the body. When there is more than one type of radiation,

the products are summed, i.e., HT � �R

wR DT,R. For space radiation, applies

only to stochastic effects and is expressed in sievert (Sv).

fluence (&): The quotient of dN by da, where dN is the number of particles

incident on a sphere of cross-sectional area da, i.e., & � dN/da. The unit

for fluence is m�2, commonly given in cm�2. In this Report, distributions

of fluence are also noted variously as a function of one or more other

variables [e.g., &(L,t), the distribution of fluence as a function of LET (L)

and time (t)].

galactic cosmic radiation (GCR): The charged-particle radiation outside

the magnetosphere, which comprise two percent electrons and positrons

and 98 percent nuclei, the latter component consisting (by fluence) of 87

percent protons, 12 percent helium ions, and 1 percent high atomic num-

ber, high-energy (HZE) particles.

gray (Gy): The special name for the unit of absorbed dose (D), kerma and

specific energy imparted, 1 Gy � 1 J kg�1.

gray equivalent (Gy-Eq): The name (NCRP, 2000a) for the unit of the

quantity gray equivalent (GT), 1 Gy-Eq � 1 J kg�1 [see also gray equiva-

lent (GT)].

gray equivalent (GT): The product of DT and Ri, where DT is the mean

absorbed dose in an organ or tissue and Ri is a recommended value for

relative biological effectiveness for deterministic effects for a given particle

type i, i.e., GT � Ri DT. An Ri value applies to the particle type incident

on the body. The dose limits for deterministic effects from space radiation

are given in the quantity GT [see also gray equivalent (Gy-Eq)].

high atomic number, high-energy (HZE) particles: Heavy ions having

an atomic number greater than that of helium (such as nitrogen, carbon,

boron, neon, argon or iron ions that are positively charged) and having

high kinetic energy.

immediate dose management: Actions taken to address the potential for

high doses from transient events in the space radiation environment that

could impact the conduct or completion of the mission or mission tasks.

Immediate dose management actions may be taken to avoid exceeding

administrative levels or dose limits.

lineal energy (y): The quotient of � by �, where � is the energy imparted

to the matter in a given volume by a single (energy deposition) event and

124 / GLOSSARY

� is the mean chord length of that volume, i.e., y � � /�. The unit for lineal

energy is J m�1, commonly given in keV �m�1.

linear energy transfer, unrestricted (L or L�): The quotient of dE by

d�, where dE is the mean energy lost by the particle, due to collisions

with electrons, in traversing a distance d�, i.e., L � dE/d�. The unit for

linear energy transfer is J m�1, commonly given in keV �m�1. Low linear

energy transfer radiations include electrons, x rays, and gamma rays.

High linear energy transfer radiations include alpha particles, heavy ions,

and interaction products of fast neutrons.

mean absorbed dose (DT): The mean absorbed dose in an organ or tissue,

obtained by integrating or averaging absorbed doses at points in the organ

or tissue.

organ dose equivalent (HT): The mean dose equivalent for an organ or

tissue, obtained by integrating or averaging dose equivalents at points in

the organ or tissue. It is the practice in the space radiation protection

community to obtain point values of absorbed dose (D) and dose equivalent

(H ) using the accepted quality factor-LET relationship [Q(L)], and then

to average the point quantities over the organ or tissue of interest by

means of computational models to obtain the organ dose equivalent (HT).

For space radiations, NCRP adopted the organ dose equivalent as an

acceptable approximation for equivalent dose (HT) for stochastic effects.

personal dosimeter: A radiation detection device worn or carried by an

individual to monitor the individual’s radiation exposure. For space activi-

ties, a device worn or carried by an astronaut in-flight.

quality factor (Q): The factor by which absorbed dose (D) at a point is

modified to obtain the dose equivalent (H) at the point, i.e., H � QD, in

order to express the effectiveness of an absorbed dose (in inducing stochas-

tic effects) on a common scale for all types of ionizing radiation. There is

a specified dependence [Q(L)] of the quality factor (Q) as a function of the

unrestricted linear energy transfer (L) in water at the point of interest.

Ri (recommended value for relative biological effectiveness): A best

estimate, for radiation protection purposes, of the relative biological effec-

tiveness for deterministic effects of a given particle type, by which the

mean absorbed dose in an organ or tissue (DT) is modified to obtain gray

equivalent (GT).

radiation safety records: The collections of information that relate to

radiation exposure of the astronauts. The radiation safety records consist

of multiple sources of information that are linked (i.e., the dosimetry

records and their supporting data and files, and other information that

bears on the radiation exposure of astronauts).

radiation weighting factor (wR): A factor used to allow for differences in

the biological effectiveness between different radiations when calculating

equivalent dose (HT) (see equivalent dose). These factors are independent

of the tissue or organ irradiated.

reentrant electrons: Decay products of unstable nuclei produced by the

nuclear interactions of trapped galactic cosmic radiation (GCR) ions.

relative biological effectiveness (RBE): A factor used to compare the

biological effectiveness of absorbed doses from different types of ionizing

GLOSSARY / 125

radiation, determined experimentally. RBE is the ratio of the absorbed

dose of a reference radiation (usually taken as 250 kVp x rays) to the

absorbed dose of the radiation in question required to produce an identical

biological effect in a particular experimental organism or tissue.

sievert (Sv): The special name for the unit of effective dose (E), equiva-

lent dose (HT), dose equivalent (H), and organ dose equivalent (HT),

1 Sv � 1 J kg�1.

solar particle event (SPE): An eruption at the sun that releases a large

number of particles (primarily protons) over the course of hours or days.

splash albedo electrons: Electrons scattered upward by interactions in

the atmosphere.

stochastic effects: Effects, the probability of occurrence which, rather

than their severity, is a function of radiation dose without threshold,

e.g., cancer.

tissue weighting factor (wT): A factor representing the ratio of risk of

stochastic effects attributable to irradiation of a given organ or tissue to

the total risk when the whole body is irradiated uniformly. The factor is

independent of the type of radiation or energy of the radiation.

Acronyms, Abbreviations,and Main Symbols

AE-8 Aerospace Electron Environment, Version 8

(a model of trapped electron radiation)

ALARA as low as reasonably achievable

AP-8 Aerospace Proton Environment, Version 8

(a model of trapped proton radiation)

CERF European Organisation for Nuclear Research/

European Union reference field

D absorbed dose (at a point)

D·

absorbed dose rate (at a point)

DT mean absorbed dose in an organ or tissue

DOELAP U.S. Department of Energy Laboratory

Accreditation Program

E effective dose

E particle energy

EVA extravehicular activity (astronaut in space suit

outside Space Shuttle or ISS)

FISH fluorescence in situ hybridization

f (y) frequency distribution of lineal energy

GT gray equivalent, when used as a quantity

GCR galactic cosmic radiation

Gy-Eq gray equivalent, when used as the name of the

unit for the quantity GT

H dose equivalent (at a point)

HT equivalent dose

HT organ dose equivalent

HRE highly-relativistic electron

HTR high temperature ratio

HZE high atomic number, high-energy

ISS International Space Station

L (or L�) unrestricted linear energy transfer

L� restricted linear energy transferLEO low-Earth orbitLET linear energy transferMir the Russian orbital space stationn�1 per nucleonNVLAP National Voluntary Laboratory Accreditation

Program

126

ACRONYMS, ABBREVIATIONS, AND MAIN SYMBOLS / 127

OSL optically stimulated luminescence

OSLD optically stimulated luminescent dosimeter

PADC/CR-39� polyallyl diglycol carbonate [trade name CR-39

(PPG Industries, Inc., Pittsburgh, Pennsylvania)]

PCC prematurely condensed chromosome

& fluence

PNTD plastic nuclear track detector

PSD position sensitive solid-state detector

Q quality factor

Q(L) quality factor, as a function of linear energy

transfer

Ri recommended value (best estimate) for relative

biological effectiveness of a given particle type i

(deterministic effects)

RBE relative biological effectiveness (experimental)

RHO radiation health officer

RPL radiophotoluminescense

SAA South Atlantic Anomaly

SPE solar particle event

SRAG Space Radiation Analysis Group

t time

TEPC tissue equivalent proportional counter

TL thermoluminescence

TLD thermoluminescent dosimeters

TLD-100� lithium fluoride, doped with magnesium and

titanium (92.5 percent 7Li, 7.5 percent 6Li)

(Bicron, Saint-Gobain Industries, Cleveland,

Ohio)

TLD-300� calcium fluoride, doped with thulium (Bicron,

Saint-Gobain Industries, Cleveland, Ohio)


titanium (enriched to 95.5 percent 6Li) (Bicron,



titanium (enriched to 99.5 percent 7Li) (Bicron,


y lineal energy

Z atomic number or particle charge (the atomic

number is numerically equal to the electrical

charge on the nucleus of a heavy ion)

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BADHWAR, G.D., SILBERBERG, R., TSAO, C.H., TOWNSEND, L.W.

and TRIPATHI, R.K. (1995). HZETRN: Description of a Free-Space Ion

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Space Information, Hanover, Maryland).

WILSON, J.W., TWEED, J., NEALY, J.E., ZEITLIN, C., KIM, M.H.Y.,

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The NCRP

The National Council on Radiation Protection and Measurements is a

nonprofit corporation chartered by Congress in 1964 to:

1. Collect, analyze, develop and disseminate in the public interest informa-

tion and recommendations about (a) protection against radiation and

(b) radiation measurements, quantities and units, particularly those con-

cerned with radiation protection.

2. Provide a means by which organizations concerned with the scientific

and related aspects of radiation protection and of radiation quantities,

units and measurements may cooperate for effective utilization of their

combined resources, and to stimulate the work of such organizations.

3. Develop basic concepts about radiation quantities, units and measure-

ments, about the application of these concepts, and about radiation

protection.

4. Cooperate with the International Commission on Radiological Protection,

the International Commission on Radiation Units and Measurements,

and other national and international organizations, governmental and

private, concerned with radiation quantities, units and measurements

and with radiation protection.

The Council is the successor to the unincorporated association of scien-

tists known as the National Committee on Radiation Protection and Mea-

surements and was formed to carry on the work begun by the Committee

in 1929.

The participants in the Council’s work are the Council members and

members of scientific and administrative committees. Council members are

selected solely on the basis of their scientific expertise and serve as individu-

als, not as representatives of any particular organization. The scientific

committees, composed of experts having detailed knowledge and competence

in the particular area of the committee’s interest, draft proposed recommen-

dations. These are then submitted to the full membership of the Council

for careful review and approval before being published.

The following comprise the current officers and membership of the

Council:

Officers

President Thomas S. TenfordeVice President Kenneth R. KaseSecretary and Treasurer William M. BecknerAssistant Secretary Michael F. McBride

141

142 / THE NCRP

Members

John F. Ahearne Ethel S. Gilbert Bruce A. NapierLarry E. Anderson Joel E. Gray Carl J. PaperielloBenjamin R. Archer Andrew J. Grosovsky Ronald C. PetersenMary M. Austin-Seymour Raymond A. Guilmette R. Julian PrestonHarold L. Beck William R. Hendee Jerome S. PuskinEleanor A. Blakely John W. Hirshfeld Marvin RosensteinJohn D. Boice, Jr. David G. Hoel Lawrence N. RothenbergThomas B. Borak F. Owen Hoffman Henry D. RoyalAndre Bouville Geoffrey R. Howe Michael T. RyanLeslie A. Braby Kenneth R. Kase Jonathan M. SametDavi J. Brenner Ann R. Kennedy Stephen M. SeltzerAntone L. Brooks David C. Kocher Roy E. ShoreJerrold T. Bushberg Ritsuko Komaki Edward A. SicklesJohn F. Cardella Amy Kronenberg David H. SlineyShih-Yew Chen Charles E. Land Paul SlovicChung-Kwang Chou Susan M. Langhorst Daniel J. StromMary E. Clark Richard W. Leggett Louise C. StrongJames E. Cleaver Howard L. Liber Thomas S. TenfordeJ. Donald Cossairt James C. Lin Lawrence W. TownsendAllen G. Croff Jill Lipoti Lois B. TravisFrancis A. Cucinotta John B. Little Robert L. UllrichPaul M. DeLuca Jay H. Lubin Richard J. VetterCarter Denniston C. Douglas Maynard Daniel WartenbergGail de Planque Claire M. Mays David A. WeberJohn F. Dicello Barbara J. McNeil F. Ward WhickerSarah S. Donaldson Fred A. Mettler, Jr. Chris G. WhippleWilliam P. Dornsife Charles W. Miller J. Frank WilsonStephen A. Feig Jack Miller Susan D. WiltshireH. Keith Florig Kenneth L. Miller Marco ZaiderKenneth R. Foster William F. Morgan Pasquale ZanzonicoJohn F. Frazier John E. Moulder Marvin C. ZiskinThomas F. Gesell David S. Myers

Honorary Members

Lauriston S. Taylor, Honorary PresidentWarren K. Sinclair, President Emeritus; Charles B. Meinhold, President Emeritus

S. James Adelstein, Honorary Vice PresidentW. Roger Ney, Executive Director Emeritus

Seymour Abrahamson Patricia W. Durbin Robert J. NelsenEdward L. Alpen Keith F. Eckerman Wesley L. NyborgLynn R. Anspaugh Thomas S. Ely John W. Poston, Sr.John A. Auxier Richard F. Foster Andrew K. PoznanskiWilliam J. Bair R.J. Michael Fry Chester R. RichmondBruce B. Boecker Robert O. Gorson William L. RussellVictor P. Bond Arthur W. Guy Eugene L. SaengerRobert L. Brent Eric J. Hall William J. SchullReynold F. Brown Naomi H. Harley J. Newell StannardMelvin C. Carter Donald G. Jacobs John B. StorerRandall S. Caswell Bernd Kahn John E. TillFrederick P. Cowan Roger O. McClellan Arthur C. UptonJames F. Crow Dade W. Moeller George L. VoelzGerald D. Dodd A. Alan Moghissi Edward W. Webster

THE NCRP / 143

Lauriston S. Taylor Lecturers

Herbert M. Parker (1977) The Squares of the Natural Numbers in RadiationProtection

Sir Edward Pochin (1978) Why be Quantitative about Radiation RiskEstimates?

Hymer L. Friedell (1979) Radiation Protection—Concepts and Trade OffsHarold O. Wyckoff (1980) From ‘‘Quantity of Radiation’’ and ‘‘Dose’’ to

‘‘Exposure’’ and ‘‘Absorbed Dose’’—An Historical ReviewJames F. Crow (1981) How Well Can We Assess Genetic Risk? Not VeryEugene L. Saenger (1982) Ethics, Trade-offs and Medical RadiationMerril Eisenbud (1983) The Human Environment—Past, Present and FutureHarald H. Rossi (1984) Limitation and Assessment in Radiation ProtectionJohn H. Harley (1985) Truth (and Beauty) in Radiation MeasurementHerman P. Schwan (1986) Biological Effects of Non-ionizing Radiations:

Cellular Properties and InteractionsSeymour Jablon (1987) How to be Quantitative about Radiation Risk

EstimatesBo Lindell (1988) How Safe is Safe Enough?Arthur C. Upton (1989) Radiobiology and Radiation Protection: The Past

Century and Prospects for the FutureJ. Newell Stannard (1990) Radiation Protection and the Internal Emitter

SagaVictor P. Bond (1991) When is a Dose Not a Dose?Edward W. Webster (1992) Dose and Risk in Diagnostic Radiology: How Big?

How Little?Warren K. Sinclair (1993) Science, Radiation Protection and the NCRPR.J. Michael Fry (1994) Mice, Myths and MenAlbrecht Kellerer (1995) Certainty and Uncertainty in Radiation ProtectionSeymour Abrahamson (1996) 70 Years of Radiation Genetics: Fruit Flies, Mice

and HumansWilliam J. Bair (1997) Radionuclides in the Body: Meeting the Challenge!Eric J. Hall (1998) From Chimney Sweeps to Astronauts: Cancer Risks in the

WorkplaceNaomi H. Harley (1999) Back to BackgroundS. James Adelstein (2000) Administered Radioactivity: Unde Venimus Quoque

ImusWesley L. Nyborg (2001) Assuring the Safety of Medical Diagnostic

UltrasoundR. Julian Preston (2002) Developing Mechanistic Data for Incorporation into

Cancer Risk Assessment: Old Problems and New Approaches

Currently, the following committees are actively engaged in formulating

recommendations:

SC 1 Basic Criteria, Epidemiology, Radiobiology and Risk

SC 1-4 Extrapolation of Risks from Non-Human Experimental

Systems to Man

SC 1-7 Information Needed to Make Radiation Protection

Recommendations for Travel Beyond Low-Earth Orbit

SC 1-8 Risk to Thyroid from Ionizing Radiation

SC 1-10 Review of Cohen’s Radon Research Methods

SC 1-11 Radiation Protection and Measurement for Neutron

Surveillance Scanners

144 / THE NCRP

SC 1-12 Exposure Limits for Security Surveillance Devices

SC 9 Structural Shielding Design and Evaluation for Medical Use of

X Rays and Gamma Rays of Energies Up to 10 MeV

SC 46 Operational Radiation Safety

SC 46-8 Radiation Protection Design Guidelines for Particle

Accelerator Facilities

SC 46-13 Design of Facilities for Medical Radiation Therapy

SC 46-16 Radiation Protection in Veterinary Medicine

SC 46-17 Radiation Protection in Educational Institutions

SC 57-15 Uranium Risk

SC 57-17 Radionuclide Dosimetry Models for Wounds

SC 64 Environmental Issues

SC 64-22 Design of Effective Effluent and Environmental

Monitoring Programs

SC 64-23 Cesium in the Environment

SC 66 Biological Effects and Exposure Criteria for Ultrasound

SC 72 Radiation Protection in Mammography

SC 85 Risk of Lung Cancer from Radon

SC 87 Radioactive and Mixed Waste

SC 87-1 Waste Avoidance and Volume Reduction

SC 87-3 Performance Assessment

SC 87-5 Risk Management Analysis for Decommissioned Sites

SC 89 Nonionizing Electromagnetic Fields

SC 89-3 Biological Effects of Extremely Low-Frequency Electric

and Magnetic Fields

SC 89-4 Biological Effects and Exposure Recommendations for

Modulated Radiofrequency Fields

SC 89-6 Wireless Telecommunications Safety Issues for Building

Owners and Managers

SC 91 Radiation Protection in Medicine

SC 91-1 Precautions in the Management of Patients Who Have

Received Therapeutic Amounts of Radionuclides

SC 91-2 Radiation Protection in Dentistry

SC 92 Public Policy and Risk Communication

SC 93 Radiation Measurement and Dosimetry

In recognition of its responsibility to facilitate and stimulate cooperation

among organizations concerned with the scientific and related aspects of

radiation protection and measurement, the Council has created a category

of NCRP Collaborating Organizations. Organizations or groups of organiza-

tions that are national or international in scope and are concerned with

scientific problems involving radiation quantities, units, measurements and

THE NCRP / 145

effects, or radiation protection may be admitted to collaborating status by

the Council. Collaborating Organizations provide a means by which the

NCRP can gain input into its activities from a wider segment of society.

At the same time, the relationships with the Collaborating Organizations

facilitate wider dissemination of information about the Council’s activities,

interests and concerns. Collaborating Organizations have the opportunity

to comment on draft reports (at the time that these are submitted to the

members of the Council). This is intended to capitalize on the fact that

Collaborating Organizations are in an excellent position to both contribute

to the identification of what needs to be treated in NCRP reports and to

identify problems that might result from proposed recommendations. The

present Collaborating Organizations with which the NCRP maintains liai-

son are as follows:

Agency for Toxic Substances and Disease Registry

American Academy of Dermatology

American Academy of Environmental Engineers

American Academy of Health Physics

American Association of Physicists in Medicine

American College of Medical Physics

American College of Nuclear Physicians

American College of Occupational and Environmental Medicine

American College of Radiology

American Dental Association

American Industrial Hygiene Association

American Institute of Ultrasound in Medicine

American Insurance Services Group

American Medical Association

American Nuclear Society

American Pharmaceutical Association

American Podiatric Medical Association

American Public Health Association

American Radium Society

American Roentgen Ray Society

American Society for Therapeutic Radiology and Oncology

American Society of Health-System Pharmacists

American Society of Radiologic Technologists

Association of University Radiologists

Bioelectromagnetics Society

Campus Radiation Safety Officers

College of American Pathologists

Conference of Radiation Control Program Directors, Inc.

Council on Radionuclides and Radiopharmaceuticals

Defense Threat Reduction Agency

Electric Power Research Institute

Federal Communications Commission

Federal Emergency Management Agency

Genetics Society of America

146 / THE NCRP

Health Physics Society

Institute of Electrical and Electronics Engineers, Inc.

Institute of Nuclear Power Operations

International Brotherhood of Electrical Workers

National Aeronautics and Space Administration

National Association of Environmental Professionals

National Electrical Manufacturers Association

National Institute for Occupational Safety and Health

National Institute of Standards and Technology

Nuclear Energy Institute

Office of Science and Technology Policy

Oil, Chemical and Atomic Workers

Radiation Research Society

Radiological Society of North America

Society for Risk Analysis

Society of Chairmen of Academic Radiology Departments

Society of Nuclear Medicine

Society of Skeletal Radiology

U.S. Air Force

U.S. Army

U.S. Coast Guard

U.S. Department of Energy

U.S. Department of Housing and Urban Development

U.S. Department of Labor

U.S. Department of Transportation

U.S. Environmental Protection Agency

U.S. Navy

U.S. Nuclear Regulatory Commission

U.S. Public Health Service

Utility Workers Union of America

The NCRP has found its relationships with these organizations to be

extremely valuable to continued progress in its program.

Another aspect of the cooperative efforts of the NCRP relates to the

Special Liaison relationships established with various governmental organi-

zations that have an interest in radiation protection and measurements.

This liaison relationship provides: (1) an opportunity for participating orga-

nizations to designate an individual to provide liaison between the organiza-

tion and the NCRP; (2) that the individual designated will receive copies of

draft NCRP reports (at the time that these are submitted to the members

of the Council) with an invitation to comment, but not vote; and (3) that new

NCRP efforts might be discussed with liaison individuals as appropriate, so

that they might have an opportunity to make suggestions on new studies

and related matters. The following organizations participate in the Special

Liaison Program:

Australian Radiation Laboratory

Bundesamt fur Strahlenschutz (Germany)

THE NCRP / 147

Canadian Nuclear Safety Commission

Central Laboratory for Radiological Protection (Poland)

China Institute for Radiation Protection

Commisariat a l’Energie Atomique

Commonwealth Scientific Instrumentation Research Organization

(Australia)

European Commission

Health Council of the Netherlands

International Commission on Non-Ionizing Radiation Protection

Japan Radiation Council

Korea Institute of Nuclear Safety

National Radiological Protection Board (United Kingdom)

Russian Scientific Commission on Radiation Protection

South African Forum for Radiation Protection

World Association of Nuclear Operations

The NCRP values highly the participation of these organizations in the

Special Liaison Program.

The Council also benefits significantly from the relationships established

pursuant to the Corporate Sponsor’s Program. The program facilitates the

interchange of information and ideas and corporate sponsors provide valu-

able fiscal support for the Council’s program. This developing program cur-

rently includes the following Corporate Sponsors:

3M Corporate Health Physics

Amersham Health

Duke Energy Corporation

ICN Biomedicals, Inc.

Landauer, Inc.

Nuclear Energy Institute

Philips Medical Systems

Southern California Edison

The Council’s activities are made possible by the voluntary contribution

of time and effort by its members and participants and the generous support

of the following organizations:

3M Health Physics Services

Agfa Corporation

Alfred P. Sloan Foundation

Alliance of American Insurers

American Academy of Dermatology

American Academy of Health Physics

American Academy of Oral and Maxillofacial Radiology

American Association of Physicists in Medicine

American Cancer Society

American College of Medical Physics

American College of Nuclear Physicians

American College of Occupational and Environmental Medicine

148 / THE NCRP

American College of Radiology

American College of Radiology Foundation

American Dental Association

American Healthcare Radiology Administrators

American Industrial Hygiene Association

American Insurance Services Group

American Medical Association

American Nuclear Society

American Osteopathic College of Radiology

American Podiatric Medical Association

American Public Health Association

American Radium Society

American Roentgen Ray Society

American Society of Radiologic Technologists

American Society for Therapeutic Radiology and Oncology

American Veterinary Medical Association

American Veterinary Radiology Society

Association of University Radiologists

Battelle Memorial Institute

Canberra Industries, Inc.

Chem Nuclear Systems

Center for Devices and Radiological Health

College of American Pathologists

Committee on Interagency Radiation Research and Policy

Coordination

Commonwealth of Pennsylvania

Consumers Power Company

Council on Radionuclides and Radiopharmaceuticals

Defense Nuclear Agency

Eastman Kodak Company

Edison Electric Institute

Edward Mallinckrodt, Jr. Foundation

EG&G Idaho, Inc.

Electric Power Research Institute

Federal Emergency Management Agency

Florida Institute of Phosphate Research

Fuji Medical Systems, U.S.A., Inc.

Genetics Society of America

Health Effects Research Foundation (Japan)

Health Physics Society

Institute of Nuclear Power Operations

James Picker Foundation

Martin Marietta Corporation

Motorola Foundation

National Aeronautics and Space Administration

National Association of Photographic Manufacturers

National Cancer Institute

National Electrical Manufacturers Association

THE NCRP / 149

National Institute of Standards and Technology

Picker International

Public Service Electric and Gas Company

Radiation Research Society

Radiological Society of North America

Richard Lounsbery Foundation

Sandia National Laboratory

Siemens Medical Systems, Inc.

Society of Nuclear Medicine

Society of Pediatric Radiology


U.S. Department of Labor

U.S. Environmental Protection Agency

U.S. Navy

U.S. Nuclear Regulatory Commission

Victoreen, Inc.

Westinghouse Electric Corporation

Initial funds for publication of NCRP reports were provided by a grant

from the James Picker Foundation.

The NCRP seeks to promulgate information and recommen-dations

based on leading scientific judgment on matters of radiation protection and

measurement and to foster cooperation among organizations concerned with

these matters. These efforts are intended to serve the public interest and

the Council welcomes comments and suggestions on its reports or activities

from those interested in its work.

NCRP Publications

Information on NCRP publications may be obtained from the NCRP

website (http://www.ncrp.com), e-mail ([email protected]), by telephone

(800-229-2652, Ext. 25), or fax (301-907-8768). The address is:

NCRP Publications

7910 Woodmont Avenue

Suite 400

Bethesda, MD 20814-3095

Abstracts of NCRP reports published since 1980, abstracts of all NCRP

commentaries, and the text of all NCRP statements are available at the

NCRP website. Currently available publications are listed below.

NCRP Reports

No. Title

8 Control and Removal of Radioactive Contamination in

Laboratories (1951)

22 Maximum Permissible Body Burdens and Maximum Permissible

Concentrations of Radionuclides in Air and in Water for

Occupational Exposure (1959) [Includes Addendum 1 issued in

August 1963]

25 Measurement of Absorbed Dose of Neutrons, and of Mixtures of

Neutrons and Gamma Rays (1961)

27 Stopping Powers for Use with Cavity Chambers (1961)

30 Safe Handling of Radioactive Materials (1964)

32 Radiation Protection in Educational Institutions (1966)

35 Dental X-Ray Protection (1970)

36 Radiation Protection in Veterinary Medicine (1970)

37 Precautions in the Management of Patients Who Have Received

Therapeutic Amounts of Radionuclides (1970)

38 Protection Against Neutron Radiation (1971)

40 Protection Against Radiation from Brachytherapy Sources (1972)

41 Specification of Gamma-Ray Brachytherapy Sources (1974)

42 Radiological Factors Affecting Decision-Making in a Nuclear

Attack (1974)

150

NCRP PUBLICATIONS / 151

44 Krypton-85 in the Atmosphere—Accumulation, Biological

Significance, and Control Technology (1975)

46 Alpha-Emitting Particles in Lungs (1975)

47 Tritium Measurement Techniques (1976)

49 Structural Shielding Design and Evaluation for Medical Use of

X Rays and Gamma Rays of Energies Up to 10 MeV (1976)

50 Environmental Radiation Measurements (1976)

52 Cesium-137 from the Environment to Man: Metabolism and Dose

(1977)

54 Medical Radiation Exposure of Pregnant and Potentially

Pregnant Women (1977)

55 Protection of the Thyroid Gland in the Event of Releases of

Radioiodine (1977)

57 Instrumentation and Monitoring Methods for Radiation

Protection (1978)

58 A Handbook of Radioactivity Measurements Procedures, 2nd ed.

(1985)

60 Physical, Chemical, and Biological Properties of Radiocerium

Relevant to Radiation Protection Guidelines (1978)

61 Radiation Safety Training Criteria for Industrial Radiography

(1978)

62 Tritium in the Environment (1979)

63 Tritium and Other Radionuclide Labeled Organic Compounds

Incorporated in Genetic Material (1979)

64 Influence of Dose and Its Distribution in Time on Dose-Response

Relationships for Low-LET Radiations (1980)

65 Management of Persons Accidentally Contaminated with

Radionuclides (1980)

67 Radiofrequency Electromagnetic Fields—Properties, Quantities

and Units, Biophysical Interaction, and Measurements (1981)

68 Radiation Protection in Pediatric Radiology (1981)

69 Dosimetry of X-Ray and Gamma-Ray Beams for Radiation

Therapy in the Energy Range 10 keV to 50 MeV (1981)

70 Nuclear Medicine—Factors Influencing the Choice and Use of

Radionuclides in Diagnosis and Therapy (1982)

72 Radiation Protection and Measurement for Low-Voltage Neutron

Generators (1983)

73 Protection in Nuclear Medicine and Ultrasound Diagnostic

Procedures in Children (1983)

74 Biological Effects of Ultrasound: Mechanisms and Clinical

Implications (1983)

75 Iodine-129: Evaluation of Releases from Nuclear Power

Generation (1983)

77 Exposures from the Uranium Series with Emphasis on Radon

and Its Daughters (1984)

78 Evaluation of Occupational and Environmental Exposures to

Radon and Radon Daughters in the United States (1984)

152 / NCRP PUBLICATIONS

79 Neutron Contamination from Medical Electron Accelerators

(1984)

80 Induction of Thyroid Cancer by Ionizing Radiation (1985)

81 Carbon-14 in the Environment (1985)

82 SI Units in Radiation Protection and Measurements (1985)

83 The Experimental Basis for Absorbed-Dose Calculations in

Medical Uses of Radionuclides (1985)

84 General Concepts for the Dosimetry of Internally Deposited

Radionuclides (1985)

85 Mammography—A User’s Guide (1986)

86 Biological Effects and Exposure Criteria for Radiofrequency

Electromagnetic Fields (1986)

87 Use of Bioassay Procedures for Assessment of Internal

Radionuclide Deposition (1987)

88 Radiation Alarms and Access Control Systems (1986)

89 Genetic Effects from Internally Deposited Radionuclides (1987)

90 Neptunium: Radiation Protection Guidelines (1988)

92 Public Radiation Exposure from Nuclear Power Generation in

the United States (1987)

93 Ionizing Radiation Exposure of the Population of the United

States (1987)

94 Exposure of the Population in the United States and Canada

from Natural Background Radiation (1987)

95 Radiation Exposure of the U.S. Population from Consumer

Products and Miscellaneous Sources (1987)

96 Comparative Carcinogenicity of Ionizing Radiation and

Chemicals (1989)

97 Measurement of Radon and Radon Daughters in Air (1988)

99 Quality Assurance for Diagnostic Imaging (1988)

100 Exposure of the U.S. Population from Diagnostic Medical

Radiation (1989)

101 Exposure of the U.S. Population from Occupational Radiation

(1989)

102 Medical X-Ray, Electron Beam and Gamma-Ray Protection for

Energies Up to 50 MeV (Equipment Design, Performance and

Use) (1989)

103 Control of Radon in Houses (1989)

104 The Relative Biological Effectiveness of Radiations of Different

Quality (1990)

105 Radiation Protection for Medical and Allied Health Personnel

(1989)

106 Limit for Exposure to ‘‘Hot Particles’’ on the Skin (1989)

107 Implementation of the Principle of As Low As Reasonably

Achievable (ALARA) for Medical and Dental Personnel (1990)108 Conceptual Basis for Calculations of Absorbed-Dose

Distributions (1991)109 Effects of Ionizing Radiation on Aquatic Organisms (1991)110 Some Aspects of Strontium Radiobiology (1991)


111 Developing Radiation Emergency Plans for Academic, Medical orIndustrial Facilities (1991)

112 Calibration of Survey Instruments Used in Radiation Protectionfor the Assessment of Ionizing Radiation Fields andRadioactive Surface Contamination (1991)

113 Exposure Criteria for Medical Diagnostic Ultrasound: I. CriteriaBased on Thermal Mechanisms (1992)

114 Maintaining Radiation Protection Records (1992)115 Risk Estimates for Radiation Protection (1993)116 Limitation of Exposure to Ionizing Radiation (1993)117 Research Needs for Radiation Protection (1993)118 Radiation Protection in the Mineral Extraction Industry (1993)119 A Practical Guide to the Determination of Human Exposure to

Radiofrequency Fields (1993)120 Dose Control at Nuclear Power Plants (1994)121 Principles and Application of Collective Dose in Radiation

Protection (1995)122 Use of Personal Monitors to Estimate Effective Dose Equivalent

and Effective Dose to Workers for External Exposure to Low-LET Radiation (1995)

123 Screening Models for Releases of Radionuclides to Atmosphere,Surface Water, and Ground (1996)

124 Sources and Magnitude of Occupational and Public Exposuresfrom Nuclear Medicine Procedures (1996)

125 Deposition, Retention and Dosimetry of Inhaled RadioactiveSubstances (1997)

126 Uncertainties in Fatal Cancer Risk Estimates Used in RadiationProtection (1997)

127 Operational Radiation Safety Program (1998)128 Radionuclide Exposure of the Embryo/Fetus (1998)129 Recommended Screening Limits for Contaminated Surface Soil

and Review of Factors Relevant to Site-Specific Studies (1999)130 Biological Effects and Exposure Limits for ‘‘Hot Particles’’ (1999)131 Scientific Basis for Evaluating the Risks to Populations from

Space Applications of Plutonium (2001)132 Radiation Protection Guidance for Activities in Low-Earth Orbit

(2000)133 Radiation Protection for Procedures Performed Outside the

Radiology Department (2000)134 Operational Radiation Safety Training (2000)135 Liver Cancer Risk from Internally-Deposited Radionuclides

(2001)

136 Evaluation of the Linear-Nonthreshold Dose-Response Model forIonizing Radiation (2001)

137 Fluence-Based and Microdosimetric Event-Based Methods forRadiation Protection in Space (2001)

138 Management of Terrorist Events Involving Radioactive Material(2001)

141 Managing Potentially Radioactive Scrap Metal (2002)

142 Operational Radiation Safety Program for Astronauts in Low-Earth Orbit: A Basic Framework (2002)


Binders for NCRP reports are available. Two sizes make it possible

to collect into small binders the ‘‘old series’’ of reports (NCRP Reports

Nos. 8-30) and into large binders the more recent publications (NCRP Reports

Nos. 32-138, 141-142). Each binder will accommodate from five to seven

reports. The binders carry the identification ‘‘NCRP Reports’’ and come with

label holders which permit the user to attach labels showing the reports

contained in each binder.

The following bound sets of NCRP reports are also available:

Volume I. NCRP Reports Nos. 8, 22

Volume II. NCRP Reports Nos. 23, 25, 27, 30

Volume III. NCRP Reports Nos. 32, 35, 36, 37

Volume IV. NCRP Reports Nos. 38, 40, 41

Volume V. NCRP Reports Nos. 42, 44, 46

Volume VI. NCRP Reports Nos. 47, 49, 50, 51

Volume VII. NCRP Reports Nos. 52, 53, 54, 55, 57

Volume VIII. NCRP Report No. 58

Volume IX. NCRP Reports Nos. 59, 60, 61, 62, 63

Volume X. NCRP Reports Nos. 64, 65, 66, 67

Volume XI. NCRP Reports Nos. 68, 69, 70, 71, 72

Volume XII. NCRP Reports Nos. 73, 74, 75, 76

Volume XIII. NCRP Reports Nos. 77, 78, 79, 80

Volume XIV. NCRP Reports Nos. 81, 82, 83, 84, 85

Volume XV. NCRP Reports Nos. 86, 87, 88, 89

Volume XVI. NCRP Reports Nos. 90, 91, 92, 93

Volume XVII. NCRP Reports Nos. 94, 95, 96, 97

Volume XVIII. NCRP Reports Nos. 98, 99, 100

Volume XIX. NCRP Reports Nos. 101, 102, 103, 104

Volume XX. NCRP Reports Nos. 105, 106, 107, 108

Volume XXI. NCRP Reports Nos. 109, 110, 111

Volume XXII. NCRP Reports Nos. 112, 113, 114

Volume XXIII. NCRP Reports Nos. 115, 116, 117, 118

Volume XXIV. NCRP Reports Nos. 119, 120, 121, 122

Volume XXV. NCRP Report No. 123I and 123II

Volume XXVI. NCRP Reports Nos. 124, 125, 126, 127

Volume XXVII. NCRP Reports Nos. 128, 129, 130

Volume XXVIII. NCRP Reports Nos. 131, 132, 133

Volume XXIX. NCRP Reports Nos. 134, 135, 136, 137

(Titles of the individual reports contained in each volume are

given above.)


NCRP Commentaries

No. Title

1 Krypton-85 in the Atmosphere—With Specific Reference to the

Public Health Significance of the Proposed Controlled Release

at Three Mile Island (1980)

4 Guidelines for the Release of Waste Water from Nuclear

Facilities with Special Reference to the Public Health

Significance of the Proposed Release of Treated Waste Waters

at Three Mile Island (1987)

5 Review of the Publication, Living Without Landfills (1989)

6 Radon Exposure of the U.S. Population—Status of the Problem

(1991)

7 Misadministration of Radioactive Material in Medicine—

Scientific Background (1991)

8 Uncertainty in NCRP Screening Models Relating to Atmospheric

Transport, Deposition and Uptake by Humans (1993)

9 Considerations Regarding the Unintended Radiation Exposure of

the Embryo, Fetus or Nursing Child (1994)

10 Advising the Public about Radiation Emergencies: A Document

for Public Comment (1994)

11 Dose Limits for Individuals Who Receive Exposure from

Radionuclide Therapy Patients (1995)

12 Radiation Exposure and High-Altitude Flight (1995)

13 An Introduction to Efficacy in Diagnostic Radiology and Nuclear

Medicine (Justification of Medical Radiation Exposure) (1995)

14 A Guide for Uncertainty Analysis in Dose and Risk Assessments

Related to Environmental Contamination (1996)

15 Evaluating the Reliability of Biokinetic and Dosimetric Models

and Parameters Used to Assess Individual Doses for Risk

Assessment Purposes (1998)

Proceedings of the Annual Meeting

No. Title

1 Perceptions of Risk, Proceedings of the Fifteenth Annual

Meeting held on March 14-15, 1979 (including Taylor Lecture

No. 3) (1980)

3 Critical Issues in Setting Radiation Dose Limits, Proceedings of

the Seventeenth Annual Meeting held on April 8-9, 1981

(including Taylor Lecture No. 5) (1982)

4 Radiation Protection and New Medical Diagnostic Approaches,

Proceedings of the Eighteenth Annual Meeting held on April

6-7, 1982 (including Taylor Lecture No. 6) (1983)


5 Environmental Radioactivity, Proceedings of the Nineteenth

Annual Meeting held on April 6-7, 1983 (including Taylor

Lecture No. 7) (1983)

6 Some Issues Important in Developing Basic Radiation Protection

Recommendations, Proceedings of the Twentieth Annual

Meeting held on April 4-5, 1984 (including Taylor Lecture No. 8)

(1985)

7 Radioactive Waste, Proceedings of the Twenty-first Annual


(1986)

8 Nonionizing Electromagnetic Radiations and Ultrasound,

Proceedings of the Twenty-second Annual Meeting held on

April 2-3, 1986 (including Taylor Lecture No. 10) (1988)

9 New Dosimetry at Hiroshima and Nagasaki and Its Implications

for Risk Estimates, Proceedings of the Twenty-third Annual


(1988)

10 Radon, Proceedings of the Twenty-fourth Annual Meeting held

on March 30-31, 1988 (including Taylor Lecture No. 12) (1989)

11 Radiation Protection Today—The NCRP at Sixty Years,

Proceedings of the Twenty-fifth Annual Meeting held on


12 Health and Ecological Implications of Radioactively

Contaminated Environments, Proceedings of the Twenty-sixth

Annual Meeting held on April 4-5, 1990 (including Taylor

Lecture No. 14) (1991)

13 Genes, Cancer and Radiation Protection, Proceedings of the

Twenty-seventh Annual Meeting held on April 3-4, 1991


14 Radiation Protection in Medicine, Proceedings of the Twenty-

eighth Annual Meeting held on April 1-2, 1992 (including

Taylor Lecture No. 16) (1993)

15 Radiation Science and Societal Decision Making, Proceedings of

the Twenty-ninth Annual Meeting held on April 7-8, 1993


16 Extremely-Low-Frequency Electromagnetic Fields: Issues in

Biological Effects and Public Health, Proceedings of the

Thirtieth Annual Meeting held on April 6-7, 1994 (not

published).

17 Environmental Dose Reconstruction and Risk Implications,

Proceedings of the Thirty-first Annual Meeting held on April


18 Implications of New Data on Radiation Cancer Risk,

Proceedings of the Thirty-second Annual Meeting held on


19 The Effects of Pre- and Postconception Exposure to Radiation,

Proceedings of the Thirty-third Annual Meeting held on April

2-3, 1997, Teratology 59, 181–317 (1999)


20 Cosmic Radiation Exposure of Airline Crews, Passengers and

Astronauts, Proceedings of the Thirty-fourth Annual Meeting

held on April 1-2, 1998, Health Phys. 79, 466–613 (2000)

21 Radiation Protection in Medicine: Contemporary Issues,

Proceedings of the Thirty-fifth Annual Meeting held on April


22 Ionizing Radiation Science and Protection in the 21st Century,

Proceedings of the Thirty-sixth Annual Meeting held on April

5-6, 2000, Health Phys. 80, 317–402 (2001)

23 Fallout from Atmospheric Nuclear Tests—Impact on Science and

Society, Proceedings of the Thirty-seventh Annual Meeting

held on April 4-5, 2001, Health Phys. 82, 573–748 (2002)

Lauriston S. Taylor Lectures

No. Title

1 The Squares of the Natural Numbers in Radiation Protection by

Herbert M. Parker (1977)

2 Why be Quantitative about Radiation Risk Estimates? by Sir

Edward Pochin (1978)

3 Radiation Protection—Concepts and Trade Offs by Hymer L.

Friedell (1979) [Available also in Perceptions of Risk, see

above]

4 From ‘‘Quantity of Radiation’’ and ‘‘Dose’’ to ‘‘Exposure’’ and

‘‘Absorbed Dose’’—An Historical Review by Harold O. Wyckoff

(1980)

5 How Well Can We Assess Genetic Risk? Not Very by James F.

Crow (1981) [Available also in Critical Issues in Setting

Radiation Dose Limits, see above]

6 Ethics, Trade-offs and Medical Radiation by Eugene L. Saenger

(1982) [Available also in Radiation Protection and New

Medical Diagnostic Approaches, see above]

7 The Human Environment—Past, Present and Future by Merril

Eisenbud (1983) [Available also in Environmental

Radioactivity, see above]

8 Limitation and Assessment in Radiation Protection by Harald H.

Rossi (1984) [Available also in Some Issues Important in

Developing Basic Radiation Protection Recommendations, see

above]

9 Truth (and Beauty) in Radiation Measurement by John H.

Harley (1985) [Available also in Radioactive Waste, see above]

10 Biological Effects of Non-ionizing Radiations: Cellular Properties

and Interactions by Herman P. Schwan (1987) [Available also

in Nonionizing Electromagnetic Radiations and Ultrasound,

see above]


11 How to be Quantitative about Radiation Risk Estimates by

Seymour Jablon (1988) [Available also in New Dosimetry at

Hiroshima and Nagasaki and its Implications for Risk

Estimates, see above]

12 How Safe is Safe Enough? by Bo Lindell (1988) [Available also

in Radon, see above]

13 Radiobiology and Radiation Protection: The Past Century and

Prospects for the Future by Arthur C. Upton (1989) [Available

also in Radiation Protection Today, see above]

14 Radiation Protection and the Internal Emitter Saga by J. Newell

Stannard (1990) [Available also in Health and Ecological

Implications of Radioactively Contaminated Environments, see

above]

15 When is a Dose Not a Dose? by Victor P. Bond (1992) [Available

also in Genes, Cancer and Radiation Protection, see above]

16 Dose and Risk in Diagnostic Radiology: How Big? How Little? by

Edward W. Webster (1992)[Available also in Radiation

Protection in Medicine, see above]

17 Science, Radiation Protection and the NCRP by Warren K.

Sinclair (1993)[Available also in Radiation Science and

Societal Decision Making, see above]

18 Mice, Myths and Men by R.J. Michael Fry (1995)

19 Certainty and Uncertainty in Radiation Research by Albrecht M.

Kellerer. Health Phys. 69, 446–453 (1995).

20 70 Years of Radiation Genetics: Fruit Flies, Mice and Humans

by Seymour Abrahamson. Health Phys. 71, 624–633 (1996).

21 Radionuclides in the Body: Meeting the Challenge by William J.

Bair. Health Phys. 73, 423–432 (1997).

22 From Chimney Sweeps to Astronauts: Cancer Risks in the Work

Place by Eric J. Hall. Health Phys. 75, 357–366 (1998).

23 Back to Background: Natural Radiation and Radioactivity

Exposed by Naomi H. Harley. Health Phys. 79, 121–128

(2000).

24 Administered Radioactivity: Unde Venimus Quoque Imus by

S. James Adelstein. Health Phys. 80, 317–324 (2001).

25 Assuring the Safety of Medical Diagnostic Ultrasound by Wesley

L. Nyborg. Health Phys. 82, 578–587 (2002)

Symposium Proceedings

No. Title

1 The Control of Exposure of the Public to Ionizing Radiation in

the Event of Accident or Attack, Proceedings of a Symposium

held April 27-29, 1981 (1982)

2 Radioactive and Mixed Waste—Risk as a Basis for Waste

Classification, Proceedings of a Symposium held November 9,

1994 (1995)


3 Acceptability of Risk from Radiation—Application to Human

Space Flight, Proceedings of a Symposium held May 29, 1996

(1997)

4 21st Century Biodosimetry: Quantifying the Past and Predicting

the Future, Proceedings of a Symposium held on February 22,

2001, Radiat. Prot. Dosim. 97, No. 1, 7–80 (2001)

NCRP Statements

No. Title

1 ‘‘Blood Counts, Statement of the National Committee on

Radiation Protection,’’ Radiology 63, 428 (1954)

2 ‘‘Statements on Maximum Permissible Dose from Television

Receivers and Maximum Permissible Dose to the Skin of the

Whole Body,’’ Am. J. Roentgenol., Radium Ther. and Nucl.

Med. 84, 152 (1960) and Radiology 75, 122 (1960)

3 X-Ray Protection Standards for Home Television Receivers,

Interim Statement of the National Council on Radiation

Protection and Measurements (1968)

4 Specification of Units of Natural Uranium and Natural

Thorium, Statement of the National Council on Radiation

Protection and Measurements (1973)

5 NCRP Statement on Dose Limit for Neutrons (1980)

6 Control of Air Emissions of Radionuclides (1984)

7 The Probability That a Particular Malignancy May Have Been

Caused by a Specified Irradiation (1992)

8 The Application of ALARA for Occupational Exposures (1999)

9 Extension of the Skin Dose Limit for Hot Particles to Other

External Sources of Skin Irradiation (2001)

Other Documents

The following documents of the NCRP were published outside of the

NCRP report, commentary and statement series:

Somatic Radiation Dose for the General Population, Report of

the Ad Hoc Committee of the National Council on Radiation

Protection and Measurements, 6 May 1959, Science, February

19, 1960, Vol. 131, No. 3399, pages 482-486

Dose Effect Modifying Factors In Radiation Protection, Report of

Subcommittee M-4 (Relative Biological Effectiveness) of the

National Council on Radiation Protection and Measurements,

Report BNL 50073 (T-471) (1967) Brookhaven National

Laboratory (National Technical Information Service

Springfield, Virginia)

Index

Absorbed dose (D) 4, 7, 122

mean absorbed dose in an

organ or tissue (DT) 7

Accreditation programs 56–57

National Voluntary Laboratory


(NVLAP) 56–57


Laboratory Accreditation

Program (DOELAP) 56–57

Active detectors 7–10, 29–30,

46–48, 53, 85–94

Cerenkov counters 93

characteristics 47

electronic personal dosimeters

10, 48

for electrons 8, 48

ionization chambers 93–94

response time 47

semiconductor detectors 91–93

solid-state detectors 47–48

tissue equivalent proportional

counters (TEPC) 8, 47, 86–91

Active real-time monitoring

84–94

local spatial variations 84

uncertainties in external

radiation 84

Administrative levels 69–70, 72,

122

Albedo neutrons 3, 38, 122

Alpha particles 108

Aluminum oxide, doped with

carbon (Al2O3:C) 9, 108–110,

113

efficiency function 108–109

glow curve 108–109

Americium-241 102, 107

Area monitoring 7, 42–43, 53

for electrons 53

recommendations 53

160

solid-state detectors 53

tissue equivalent proportional

counters (TEPC) 53

As low as reasonably achievable

(ALARA) 1, 2, 7, 11–12, 23,

68–69, 70–73, 122

continuous in-flight review

71–72

definition 68

examples 12

graded approach to

implementation 72

postflight review 70, 72

preflight planning 71

radiation safety training 73

role of radiation health officer

(RHO) 70

spacecraft and space suit

design 70–71

Astronauts 1, 2, 13, 15, 26, 71,

73–74

role in management of dose 26

wearing of personal dosimeters

13, 71, 73

Aviation activities 68, 80

estimating dose 68

Background radiation 68

Biochemical indicators 60

Biodosimetry 11, 59–65, 80

biochemical indicators 60

changes in cellular markers 65

chromosomal aberrations 59

cytogenetic alterations 61–63

definition and purpose 59

electron spin resonance 59–60

erythrocytes with transferrin

receptors 60

fluorescence in situ

hybridization (FISH) 11, 64

future considerations 11, 65

INDEX / 161

gene mutation assays 60–61

gene mutations 59

molecular profiling 65

recommendations 64–65

summary for methods 63

use in space program 63–64

Biomedical research 4, 26, 68, 80

expert radiation panel 26

informed consent 26

Institutional Review Board 26

internal radiation 80

Boltzmann equation 95

Bone marrow 31, 80

Bragg-Gray cavity theory 86

Bremsstrahlung 30, 85

Carbon-12 89, 102

Calcium fluoride, doped with

thulium (CaF2:Tm) 105–108,

113

annealing treatments 105



peak-height ratio 105–107

Calibration of dose measurements

10, 57

angle dependence 10

conditions 57

definition 57

determination of response

characteristics 57

particle energy ranges 10

response characteristics 10, 57

Cancer mortality risk coefficients

20

Cataracts 20

Cellulose nitrate 120

Cerenkov counters 93

design and operation 93

Cesium-137 49–50, 55, 100–101,

112

Charged ions 40, 45

nuclear fragments 45

radiation transport codes 40

Chromosomal aberrations 62–63

application to astronauts 62–63

application to radiation

accidents 62

atomic-bomb survivors 62

fluorescence in situ

hybridization (FISH) 62–63

lower level of detection 63

occupationally and medically

exposed persons 62

Cobalt-60 55, 100–102, 104, 108,

110

Computational methods 95–98

basic quantum mechanics 98

Boltzmann equation 95

collision with atomic electrons

95–96

cross sections 95–97

elastic nuclear scattering 95–96

for space applications 96–97

Monte Carlo 98

nuclear reactive processes

95–96

Pauli principle 98

Computerized anatomical models

41

female 41

male 41

Conversion coefficients 10–11, 52,

57–58

in calibration 57–58

Coronal mass ejection 122

Cesium-137 110

Cytogenetic alterations 61–63

chromosomal aberrations 62–63

cytochalasin B technique 61

micronuclei 61

prematurely condensed

chromosomes 62–63

Delta rays 88–90

Deterministic effects 4, 18, 20,

122

Diagnostic and therapeutic

medical procedures 68, 79, 81

linkage to dosimetry records 79

Direct ion storage dosimeters 51

design and characteristics 51

Dose assessment 6, 35–43, 59

approach for space radiation

35, 37

biodosimetry 59

162 / INDEX

exterior exposure field 38

interior exposure field 39

tissue exposure fields 41

Dose equivalent (H) 4, 7, 20, 35,

42, 55, 112–114, 122

calculation procedure 35

determination in combined

high- and low-LET fields

113–114

estimation from

thermoluminescent dosimeter

(TLD) data 112–114

limitations in use of

thermoluminescent

dosimeters (TLD) with

neutrons 113

Dose limits 3–4, 17–19, 21,

68–70, 72, 122

bone marrow 18, 21

deterministic effects 3, 21

lens of the eye 18, 21

skin 18, 21

sources of exposure included 68

stochastic effects 3

ten-year career limits 21

Dose measurements 55–58

accuracy 55–57

calibration 55–58

dependence of response on

energy and angle of incidence

57

general requirements 56



(NVLAP) 57

performance testing 55–56

recommendations of ICRP and

ICRU 55–56

response characteristics 58

uncertainty of laboratory

calibrations 58



Program (DOELAP) 57

Dosimetry measurements 42–45

radiation response

characteristics 45

time considerations 45

Dosimetry records 13–14,

122–123

annual confidential record 13

cumulative dose from space

flight 13

effective dose (E) 13

gray equivalent (GT) 13

mission-related aviation

activities 13

mission-related biomedical

research 13

other information linked to

dosimetry record 13

purposes 13

updates of record 13–14

Earth’s geomagnetic field 38

Effective dose (E) 4, 7, 13, 18–20,

22, 36, 52, 56, 80–81, 123

approach for space radiation

20, 22


direct estimation with

thermoluminescent

dosimeters (TLD) and plastic

nuclear track detectors

(PNTD) 52

records 80–81

uncertainty in estimation 56

Efficiency function 103–105,

107–109, 110–114

Electron detectors 48–49

ionization chamber 49

Electron spin resonance 59

Electronic personal dosimeters

48, 55


direct ion storage dosimeters 48

for low-LET radiation 55

Electrons 5, 10, 30, 38-40, 58, 112

energy ranges for calibration of

measurements 58

radiation transport codes 38, 40

reentrant electrons 5

splash albedo electrons 5

trapped electrons 5

Environmental models 6

INDEX / 163

Equivalent dose (HT) 4, 19–20,

22, 123

Equivalent gamma-ray dose

110–111

Erythrocytes with transferrin

receptors 60

Exterior exposure field 38–39

radiation environment models

38

spacecraft external

measurements 39

trapped radiation models 38

Extravehicular activity (EVA) 5,

16, 41

Extravehicular activity (EVA)

suit 31, 33

design and materials 33

Flight director 2, 26, 74



Flight rules 26, 70, 74

Flight Techniques Board 26

Flight surgeon 2, 26, 75

Radiation Integrated Product

Team 26



Fluorescence in situ hybridization

(FISH) 11, 61–65

analysis of prematurely

condensed chromosomes 11

calibration curves 11

use in space program 64

Fluorine-19 102

Galactic cosmic radiation (GCR)

3, 5, 16, 32, 38, 53, 123

Gene mutation assays 60–61

glycophorin A 61

Gray equivalent (GT) 3, 7, 13,

18–19, 36, 80–81, 123


records 80–81

Helium 108

High atomic number, high-energy

(HZE) ions 5, 10, 20, 58, 123

argon 20

carbon 10


measurements 58

helium 10

iron 10, 20

silicon 10

High transient doses 12, 17, 44,

74–76

Immediate dose management 7,

11–12, 68–72, 123

continuous in-flight review

71–72

definition 68

examples 12

flight rules 12

graded approach to

implementation 72

high transient exposures 11

in-flight radiation protection

actions 69

postflight review 72

preflight planning 71

spacecraft and space suit

design 70–71

Institutional Review Board 69

Interior exposure field 39–40, 43

shielding models 39

spacecraft internal

measurements 40

Internal radiation 80–81

biomedical research 80–81

computation of effective dose

(E) 81

records 80–81

International Commission on

Radiation Units (ICRU) 56

International Commission on

Radiological Protection (ICRP)

56

Ionization chamber 8, 53, 93–94

design and operation 93–94

for electrons 53

Ionization current 39, 49

Iron-56 89

Lens of the eye 80, 85

164 / INDEX

Lineal energy (y) 7, 40, 45, 47,

87, 123–124

Linear energy transfer (LET) 30,

45, 48, 89, 92, 115–116, 124

restricted (L�) 89, 115–116

unrestricted (L�) 30, 45, 48, 92,

124

Lithium fluoride, doped with

magnesium and titanium

(LiF:Mg,Ti) 9, 99–105,

107–109, 112–113

annealing 99–100

cross section for neutron

interaction 99

deconvoluting glow curve into

component parts 100–101

dependence on particle energy

and charge 105

dose response to high-LET

radiation 100–103

dose response to low-LET

radiation 100–103



high temperature ratio

102–103

isotopic abundance of lithium

99

response to high-energy protons

102–104

Lithium fluoride, doped with

magnesium, copper and

phosphorus (LIF:Mg,Cu,P) 9,

107–109, 113


Low-Earth orbit 5, 15–16, 29–34

albedo electrons 34

charged target fragments 32,

34

coronal mass ejections 30

Earth’s magnetic field 29

galactic cosmic radiation (GCR)

32, 34

high-energy secondary

fragments 32

high-inclination flights 29

high-inclination orbit 16

light nuclear particles 5, 31–32,

34

low-inclination flights 29

neutrons 33–34

pions 32

protons 34

radiation environment 29–30

radiation measurements 33–34

reentrant electrons 31, 34

secondary charged fragments

34

solar cycle 29

solar protons 31–32

solar wind 29

South Atlantic Anomaly (SAA)

29

splash albedo electrons 31

summary for particle types 33

trapped electrons 30, 34

trapped radiation belts 30

Mean absorbed dose in an organ

or tissue (DT) 6, 18–19, 35, 42,

124


Mean, or effective, linear energy

transfer 111–112

Microdosimetry 87

Mir Space Station 15, 52

Molecular profiling 65

Muons 112



(NVLAP) 56–57

Neon 108

Neutrons 5, 10, 20, 33, 40, 44, 58,

84, 102

albedo neutrons 5, 33, 84


measurements 58

fission spectrum neutrons 20


secondary particles 33

Occupancy factors 7, 41

Occupational dose 5

INDEX / 165

Occupational Safety and Health

Administration (OSHA) 24

Operational radiation monitoring

7

area monitors 7

personal dosimeters 7

Optically stimulated

luminescence dosimeter (OSLD)

8–9, 50, 54, 110

Al2O3:C 9, 50, 54

dependence on LET 54

determination of low-LET

component 50

Orbital precession 85

Organ dose equivalent (HT) 4, 7,

19, 22, 36, 124


Oxygen-16 89

Particle spectrometers 9, 39–40,

44, 46, 48, 54, 92

Passive devices 10, 49–52, 54, 55

direct estimate of effective dose

(E) 52

direct ion storage dosimeters 51

electronic dosimeters 49

limitations 49

optically stimulated

luminescent dosimeter

(OSLD) 49, 50

plastic nuclear track detector

(PNTD) 49, 51–52,

radiophotoluminescence (RPL)

glass dosimeters 50

superheated drop/bubble

dosimeters 52

thermoluminescent dosimeters

(TLD) 49, 50, 51

uses 49

Peak-height-ratio method 50

Personal dosimeters 6, 8, 10, 40,

43–44, 53, 54–55, 124

onboard readout 10

optically stimulated

luminescent dosimeters

(OSLD) 54

plastic nuclear track detectors

(PNTD) 44, 54

readout capabilities 55

recommendations 54–55

thermoluminescent dosimeters

(TLD) 44, 54

Photons 40, 44, 112


Plastic nuclear track detectors

(PNTD) 9, 44, 46, 51–52, 54,

113, 115–121

cellulose nitrate 51


angle of incidence 46, 52

description of method 115–116

design and characteristics

51–52

determination of high-LET

component 51–52

dosimetry of cosmic radiation

fields 117–121

high-energy protons 121

HZE particles 121

neutron dosimetry 116–117

particle tracks 115–116

response to high-LET particles

46

polyallyl diglycol carbonate

(PADC) 9, 51

polycarbonate 51

proton response 116

Pions 32

Polyallyl diglycol carbonate

(PADC) 9, 113, 116–121

calibration 117

comparison of LET spectrum

and tissue equivalent

proportional counters (TEPC)

119


angle of incidence 118

determination of dose

equivalent (H) 118

electrochemical etching

116–117

energy dependence 116

LET response 117

LET threshold 117

neutron dose equivalent (H)

response 120–121

166 / INDEX

personal dosimetry systems

120–121

proton response 116

Polycarbonate 120

Polyethylene (CH2) 71

Positrons 5

Prematurely condensed

chromosomes (PCC) 62, 64–65

Previous occupational exposure

81

Protons 5, 10, 31, 38, 44, 58, 101,

108, 112


measurements 58

target fragments 31

trapped radiation models 38

Quality assurance 57

Quality factor (Q) 4, 20, 22, 55,

124

as a function of linear energy

transfer 22

Radiation belt 3

Radiation detriment 21

cancer mortality 21

heritable effects 21

nonfatal cancer 21

Radiation environment 5

Radiation Health Officer (RHO)

2, 26


Radiation protection principles 67

applied to low-Earth orbit

(LEO) missions 67

as low as reasonably achievable

(ALARA) 67

individual dose limits 67

justification 67

Radiation-related health study 23

Radiation safety program 1–3, 15,

22–25, 66–76

administrative levels 24–25

as low as reasonably achievable

(ALARA) 24–25, 66–76

astronauts 24, 73–74

dose limit 24

extravehicular activity (EVA)

25

flight director 24–25

flight rules 24–25

flight surgeon 24–25

immediate dose management

66–76

lines of authority 67

management 24, 66–76

objectives 15, 23

organizational structure 67

preplanning 69


24–25

radiation safety training 72–76

Space Radiation Analysis

Group (SRAG) 24–25

Radiation safety records 77–83,

124

annual report to astronaut 81

content of records 78–80

continuity over time 82

definition 77

dosimetry records 77

existing records 82

files linked to the dosimetry

record 79–80

future adjustment of records

82–83

maintenance 78

missing information 79

objectives 77

prospective and retrospective

studies 81–82

records of career doses 80

retention 83

unanalyzed data files 82

updates of records 80

Radiation safety training 12–13,

72–76

astronaut 12

flight director 12, 74

flight surgeon 12, 75

other individuals 12


12, 75

radiation safety professionals

12

INDEX / 167

radiation safety support groups

75

supporting specialists 76

wearing of personal dosimeters

13, 71, 73

Radiation transport calculations

6, 42–45, 47, 95, 97

anthropomorphic models 6

radiation transport code,

HZETRN 97

radiation transport equations

95

radiation transport models 45,

47

Radiation weighting factor (wR)

18, 22, 124

Radiophotoluminescence (RPL)

glasses 50

Recommended value of the

relative biological effectiveness

(Ri) 18

Record keeping 7

Reentrant electrons 31, 124

Refractive index 93

Relative biological effectiveness

(RBE) 20, 36, 87, 124–125

Relativistic electrons 16–17

Restricted linear energy transfer

(L�) 89, 115–116

Retrospective dosimetry 23

Secondary particles 44–45

Semiconductor detectors 91–93

as a personal dosimeter 92–93

design and operation 91–92

particle telescopes 92

position sensitive solid-state

detectors 92

Shielding models 6, 39

extravehicular activity (EVA)

suits 6, 39

International Space Station

(ISS) 6, 39

Space Shuttle 6, 39

Silicon-28 89

Skin 80

Solar cycle 16–17

Solar particle event (SPE) 3, 16,

85, 125

Solid-state detectors 8, 47–48, 53

design and characteristics

47–48

Sources of exposure 4, 5

background radiation 5

biomedical research 4

diagnostic and therapeutic

medical procedures 5

occupational 5

South Atlantic Anomaly (SAA)

16, 47, 53

Space Radiation Analysis Group

(SRAG) 2, 27

implementation of as low as

reasonable achievable

(ALARA) 27


Space-related biomedical research

23

Splash albedo electrons 31, 125

Stochastic effects 4, 18, 125

lifetime excess risk of cancer 18

Stopping power ratios 86

Superheated drop/bubble

dosimeters 52


Ten-year career limits 21

age at exposure 21

female 21

male 21

other than ten-year career

length 21

Thermoluminescent dosimeters

(TLD) 8–9, 33, 41, 44, 46,

50–51, 54–55, 99–114

Al2O3:C 108

CaF2:Tm 9, 50, 51, 54, 99

correction of response for high-

LET radiation 54


angle of incidence 46

dependence on LET 54

dependency on particle type

and energy 46

168 / INDEX

determination of low-LET

component 50

discrimination between high-

and low-LET particles 113

evaluation of absorbed dose (D)

and dose equivalent (H)

110–114

glow curves 111–112

high temperature ratio 111

LiF:Mg,Cu,P 9, 50, 54, 107–108

LiF:Mg,Ti 9, 50, 54, 99–105

peak-height-ratio method 50

uncertainty in absorbed dose

(D) estimate 112

variation in efficiency 50

Tissue equivalent proportional

counters (TEPC) 7, 44–45, 47,

53–55, 86–91, 112

charged-particle equilibrium 90


distributions of lineal energy

(y) 89

dosimetry of mixed fields 87

estimation of absorbed dose (D)

90

estimation of dose equivalent

(H) 90

estimation of LET 91

model for HZE particles 89–90

operation 86

regulation of gas 86

response for constant LET 88

sensitivity to neutrons 91

simulation of small tissue

volumes 86–87

Tissue exposure fields 41

human shielding models 41

occupancy factors 41

Tissue weighting factor (wT) 4,

19–20, 125

Trapped electrons 30, 85

Trapped protons 84

Unrestricted linear energy

transfer (L�) 30, 45, 48, 92, 124



Program (DOELAP) 56–57